ASSESSMENT OF RADIOACTIVITY IN MAN

Half scan

INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA,1964 Symposium on the Assessment of Rad ÁN: 100872 v .1 c.3 Seib. UN: 612.014 S9894

000DG7b-fll4b ASSESSMENT . OF RADIOACTIVITY IN MAN

PROCEEDINGS SERIES

ASSESSMENT

OF RADIOACTIVITY IN MAN

PROCEEDINGS OF THE SYMPOSIUM ON THE ASSESSMENT OF RADIOACTIVE BODY BURDENS ' , IN M A N HELD BY THE INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR ORGANISATION AND WORLD HEALTH ORGANIZATION AT HEIDELBERG, 11-16 M Á Y 1964

In two volumes VOL.I

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1964 Symposium on the Assessment of Radioactive Body Burdens in Man, Heidelberg, 11-16 May 1964. Assessment of radioactivity in man. Proceedings . . . held by the International Atomic Energy Agency . . ..Vienna, the Agency, 1964. ■ 2 vols. (IAEA, Proceedings series)

612.014.483 616.073.75

ASSESSMENT OF RADIOACTIVE BODY BURDENS IN MAN, IAEA, VIENNA, 1964 STI/PUB/84

Printed by the IAEA in Austria October 1964 FOREWORD

This Symposium on the Assessment of Radioactive Body Burdens in Man was organized jointly by the International Atomic Energy Agency, the Inter­ national Labour Organisation and the World Health Organization and was held in Heidelberg from 11-15 J\Iay 1964. It was attended by 181 participants from 28 countries and 6 international organizations. It was the objective of the Symposium to bring together experts from the various scientific disciplines of physics, chemistry, biology, medicine and mathematics, and to survey their experience in the assessment of radio­ active body burdens in man and the resultant radiation doses. In most in­ vestigations of internal contamination the errors in the physical measure­ ments are smaller than the errors associated with the interpretation of measurements. For this reason special emphasis was laid in this meeting on the interpretation of measured data. The 67 papers and the discussions which they stimulated are published in these Proceedings produced in two volumes. Volume I includes all papers which deal with problems generally common to many isotopes: in- vivo counting, bioassay techniques, sample counting and analysis of data. Volume II includes those papers concerned with radioisotopes of specific elements: caesium, radium, radon, strontium, tritium, thorium, uranium, plutonium and rare earth elements. These Proceedings should prove invaluable to all radiation protection services entrusted with the physical surveillance of internal radiation ex­ posure of man. They should complement the studies of the International Commission on Radiological Protection (ICRP) and assist the work of the Organizations that jointly organized the meeting. The three Organizations wish to express their appreciation to the Govern­ ment of the Federal Republic of Germany for its generous invitation and to the scientists who contributed the valuable new information. EDITORIAL NOTE

The papers and discussions incorporated in the proceedings published by the International Atomic Energy Agency are edited by the Agency's edi­ torial staff to the extent considered necessary for the reader's assistance. The views expressed and the general style adopted remain, however, the responsibility of the named authors or participants. ■ For the sake of speed of publication the present Proceedings have been printed by composition typing and photo-offset lithography. Within the lim i­ tations imposed by this method, every effort has been made to maintain a high editorial standard; in particular, the units and symbols employed are to the fullest practicable extent those standardized or recommended by the competent international scientific bodies. The affiliations of authors are those given at the time of nomination. The use in these Proceedings of particular designations of countries or territories does not imply any judgement by the Agency as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of specific companies or of their products or brimd-names does not imply any endorsement or recommendation on the part of the Inter­ national Atomic Energy Agency. CONTENTS OF VOL. I

INTRODUCTION .

Formulation of relationships between the radiation exposure of tissues and the excretion rate of nuclides (SM-52/81) ...... 3 E. E. Pochin On the development of devices for the determination of total-body radioactivity in man: a historical and critical review (SM-52/58). . 15 B. Rajewsky, A. Kaul and J. Heyder

IN VIVO COUNTING

A two-crystal scanning-bed counter for accurate determination of whole-body activity (SM-52/49)...... 55 Y. Naversten A new technique for determining the distribution of radium and thorium in living persons (SM-52/36) ...... 67 С. E. M iller Shapes of scintillation spectra (SM-52/46) ...... 79 K. G. McNeill and V. K. Mohindra Human beta bremsstrahlung detection by means of thin and thick . sodium iodide crystals (S M -52/47 )...... '91 L . G. Bengtsson An improved chest phantom for studies of plutonium and americium in human lungs (SM-52/17) ...... ; ...... 115 R. G. Speight, C. O. Peabody and D. Ramsden Nouveau compteur proportionnel destiné à la détection in vivo de traces de plutonium dans les poumons (SM-52/26) ...... 131 A. La n isa rt et J. - P . M o ru c c i ' ■ Performance of an arrangement of several large-area proportional counters for the assessment of Pu239 lung burdens (SM-52/56) .... 141 R. E h ret, H. K ie fe r, R. M aushart and G. M oh rle

BIOASSAY

Sampling and analysis for assessment of body burdens (SM-52/31)... 155 J. H. H arley A survey of the methods used in the United Kingdom Atomic Energy Authority for the determination of radionuclides in urine (SM- 52/ 1 6 )...... 169 S. Jackson and N. A. Taylor Recent radiochemical procedures for bio-assay studies at Trombay (S M -52/67) ...... : ...... 195 P. R. Kamath, I. S. Bhat, Kamala Rudran, M. A. R. ,Iyengar, . Elizabeth Koshy, U rm ila S. Waingankar and Vasanti S. Khanolkai• Comparison of excretion analysis with whole-body counting for assessment of internal radioactive contaminants (SM-52/39)...... 217 C. W. S ill, J. I. Anderson and D. R. P e rc iv a l The ro le of faecal analysis in a bioassay program m e (S M -52/10) .... 231 J. D. Eakins and A. M organ - Radiochemical determination of plutonium for radiological purposes (SM- 52/3 7) ...... : ...... 245 J. M. Nielsen and T. M. Beasley A procedure for the determination of alpha-emitting plutonium in urine using a solid-state counter (SM-52/9)...... 261 F. J. Sandalls and A. M organ . A dou ble-filter device to m easure radon and thoron in the breath . (SM- 52/53) ...... 275 W. Jacobi Total counting and spectroscopy in the assessment of alpha radios activity in human tissues (S M -5 2 / 2 )...... 291 W. V. M ayneord and C. R. H ill The use of gamma-ray scintillation spectrometry in bioassay (SM- 5 2 / 8 )...... Г ...... 311 A. H olm es

INTERPRETATION OF DATA

Interpretation of bioassay data (SM-52/89) ...... 329 G. W. Dolphin and S. Jackson Influence of aerosol properties and the respiratory pattern upon hazards evaluation following inhalation exposure (SM-52/40)...... 355 R. G. Thomas A critical survey of the analysis of microscopic distribution of some bone-seeking radionuclides and assessment of absorbed dose (S M -52/88) ...... 369 W. S. S. Jee

List of Chairmen of Sessions and Secretariat of the Symposium ...... 395 INTRODUCTION (Session 1)

FORMULATION OF RELATIONSHIPS BETWEEN THE RADIATION EXPOSURE OF TISSUES . AND THE EXCRETION RATE OF NUCLIDES

' E. E. FOCHIN MEDICAL RESEARCH COUNCIL DEPARTMENT OF CLINICAL RESEARCH, UNIVERSITY COLLEGE HOSPITAL MEDICAL SCHOOL, LONDON, ENGLAND

Abstract — Résumé — Аннотация — Resumen

FORMULATION OF RELATIONSHIPS BETWEEN THE RADIATION EXPOSURE OF TISSUES AND THE EXCRETION RATE OF NUCLIDES. The organization of protection against undue occupational exposure to internal radiation involves several processes: 1 1. Decisions as to the highest dose-rates, for the body or for particular'organs-, that can be regarded as permissible; 2. Estimation, for all relevant nuclides, of the intakes, and of the body burdens, which would cause any such dose-rates to be reached or sustained; ' 3. Monitoring of exposed individuals to determine what fraction of a permissible body burden of any nuclide is retained in the body. Techniques of monitoring by whole-body counting, or by data on excretion or exhalation rates, are relevant to the Symposium. Several major problems are involved: (a) For most radionuclides, insufficient metabolic data are available to link tissue dose-rates either with body burdens or with excretion rates, at least as based on adequately large numbers of normal human subjects; (b) The large variability between different subjects in physiological functions, such as excretion or clearance rates, limits severely the inferences that can be made from isolated human observations, and restricts also the deductions as to body burden that can be drawn from the excretion rates observed in any individual; (c) The complex variation of excretion rate with time after a single intake cf many, nuclides prevents any direct deduction of body burden from excretion rate when the time course of intake is unknown, except in the special case of a nuclide which is excreted at a rate which decreases (mono-) exponentially with time since intake; (d) In particular, many radionuclides which are of particular importance because they are highly concen­ trated and long retained in certain organs, may show initial rapid excretion of the fraction of intake which is not so concentrated, and much slower and more prolonged excretion of the retained part. A small recent intake will thus contribute preponderating to the excretion as compared with that from a much larger and more important burden already localized in the critical organ; . (e) The formulations of the International Commission on Radiological Protection (ICRP), which are commonly based on simple exponential models as approximations or in the absence of more detailed metabolic information, may be adequate for estimating the dose commitments attributable* to particular intakes or body burdens, but may be much less appropriate for relating excretion rate to body burden in any individual in the case of nuclides which mix in, and are excreted from, multicompartment systems in the body, and particularly in the case of excretion rates observed after recent intake. These situations require special review in the light of evidence as to the earliest phases of turnover.

FORMULES EXPRIMANT LA RELATION ENTRE LA RADIOEXPOSITION DES TISSUS ET LE TAUX D'EX­ CRÉTION DES RADIONUCLÉIDES. L'organisation de la protection contre toute exposition professionnelle excessive à une contamination interne comporte plusieurs opérations: 1. Détermination des débits de dose maximums pouvant être considérés comme admissibles pour le corps humain ou un organe particulier; 2. Evaluation, pour tous les radionucléides qui présentent un danger à cet égard, des quantités absorbées et des charges corporelles pouvant être la cause de débits de dose de cet ordre; 3. Contrôle dosimétrique des personnes exposées en vue de déterminer quelle fraction de la charge corporelle admissible pour un nucléide quelconque est retenue dans le corps. Méthodes de contrôle au moyen d’un anthropogammamètre ou par analyse des données sur les taux d'excrétion ou d'exhalation.

3 4 E. E. POCHIN

Plusieurs problèmes importants se posent: - a) Le métabolisme de la plupart des radionucléides est insuffisamment connu pour permettre d'établir une relation entre, d’une part, les débits de dose aux tissus et, d'autre part, les charges corporelles ou les taux d’excrétion; de toute façon, les données disponibles ne sont pas fondées sur l’observation d'un nombre suffisamment élevé de sujets normaux. b) Les variations importantes qu'accusent, selon les sujets, les fonctions physiologiques, telles que les taux . d'élimination par excrétion ou par une autre voie, diminuent sensiblement la valeur des conclusions que l'on peut tirer de l'observation d'individus isolés, et celle des déductions relatives à la charge corporelle que l'on peut faire en se fondant sur les taux d'excrétion observés chez n’importe quel sujet. c) Pour de nombreux nucléides, les variations complexes du taux d’excrétion en fonction du temps après une seule absorption empêch'e de déduire directement la charge corporelle à partir du taux d'excrétion lorsqu'on ignore le régime de l'absorption dans le temps, sauf dans le cas particulier d'un nucléide excrété à une vitesse qui décroît de façon (mono-)exponentielle avec le temps depuis le moment de l’absorption. ' d) De nombreux radionucléides, qui revêtent une importance particulière en raison de leur forte concen­ tration et de leur rétention prolongée dans certains organes, peuvent être excrétés à un rythme d'abord rapide, pour la fraction moins con cen trée, puis sensiblem ent plus lent, pour la partie retenue. Il s'ensuit qu’une petite quantité absorbée à une date récente se retrouvera dans la quantité excrétée en proportion sensiblement plus importante qu’une charge beaucoup plus grande qui serait déjà fixée dans l’organe critiq u e . e) il se peut que les formules que la CIPR a établies, en se servant habituellement de modèles exponentiels simples comme approximations ou faute de renseignements plus précis sur le métabolisme, suffisent pour évaluer les fractions de dose attribuables à des absorptions ou charges corporelles pàrticulières, mais il se peut aussi qu’elles se révèlent beaucoup moins satisfaisantes si l'on veut établir une relation entre le taux d’excrétion et la charge corporelle d’un sujet dans le cas de nucléides qui se mélangent dans des systèmes à compartiments multiples du corps humain ou en sont éliminés, notamment dans le cas où les taux d'excrétion sont observés après une absorption récente. Ces cas exigent un examen . spécial à la lumière des renseignements disponibles sur les premières phases du renouvellement.

ОПРЕДЕЛЕНИЕ ЗАВИСИМОСТИ МЕЖДУ ОБЛУЧЕНИЕМ ТКАНЕЙ И СКОРОСТЬЮ ВЫ­ ВЕДЕНИЯ РАДИОИЗОТОПОВ. Организация защиты от чрезмерного профессионального вну­ треннего облучения включает несколько процессов: . 1. Принятие решений в отношении определения наивысших мощностей доз для всего организма или для отдельных органов, которые могут рассматриваться как допустимые. 2. Оценка (для всех рассматриваемых радиоизотопов) тех уровней поглощения и того содер­ жания в организме, которые могут привести к тому, что любые такие мощности дозы будут достигнуты и будут поддерживаться. 3. Дозиметрический контроль облученных лиц, чтобы определить, какая часть допустимого количества любого изотопа задерживается в организме. Темой симпозиума являются методы дозиметрического контроля посредством измерения активности всего организма или посред­ ством получения данных о скорости выделения или выдыхания. . . Возникает несколько крупных проблем: . , • . а) Для большинства радиоизотопов существует недостаточно обменных данных, чтобы связать мощности тканевых доз либо с содержанием радиоизотопов в организме,;либо со скоростями выведения, и по крайней мере таких данных, которые получены на достаточно большом количестве здоровых людей. б) Большое разнообразие в физиологических функциях у различных людей, таких, как скорости выделения или очищения, резко ограничивает те выводы, которые можно сделать из наблюдений у отдельных людей, а также ограничивает те заключения в отношении содержа­ ния изотопа в организме, которые можно сделать из скорости выделений, наблюдаемой у какого-либо одного лица. • в) Сложные вариации скорости выведения с течением времени после однократного по­ глощения для многих радиоизотопов не позволяет сделать какие-либо определенные заклю­ чения о содержании в организме радиоизотопов на основании скорости выведения, когда не и з в е с т н о , в течение какого времени произошло поглощение, за исключением особых случаев с радиоизотопом, который выделяется со скоростью, уменьшающейся (моно-) экспоненциально со временем после поглощения. ' г) В частности, многие радиоизотопы, особо важные в силу высокой концентрации и длительной задержки в некоторых органах, обнаруживают первоначальное быстрое выделение TISSUE EXPOSURE AND EXCRETION RATE 5

той фракции поглощенного количества, которая не обладает такой большой концентрацией и более медленное и более продолжительное выделение задерживающейся части. Незначи­ тельное недавнее поглощение, таким образом, окажет сильное влияние на это выделение по сравнению с выделением из большего по объему и более важного количества, уже сосредо­ точенного в критическом органе. д) Правила Международной комиссии по защите от радиоактивного излучения, которые обычно формулируются на основании простых экспоненциальных моделей, имеющих приближен­ ные значения, или при отсутствии более детальной информации относительно обмена радио­ изотопов, могут оказаться достаточными для определения величины дозы, связанных с осо­ быми видами поглощения или содержанием радиоактивных веществ в организме, но они могут оказаться менее подходящими для определения скоростей выделения изотопов у какого-либо лица в случаях с радиоизотопами, которые содержатся в многоперегородочных системах че­ ловеческого организма или выделяются из них,-и особенно в случаях скорости выделения, наблюдаемой после недавнего поглощения. Такие ситуации требуют специального исследо­ вания в свете данных, относящихся, к более ранним фазам кругооборота.

FORMULACIÓN DE RELACIONES DE DEPENDENCIA ENTRE LA RADIO EXPOSICIÓN DE LOS TEJIDOS Y LA VELOCIDAD DE EXCRECIÓN DE LOS NÚCLIDOS. La organización de un sistema de protección contra una excesiva exposición a la irradiación interna por motivos profesionales supone una serie de operaciones, a saber: • ■ 1) La adopción de dicisiones acerca de la intensidad de dosis máxima para el organismo entero o para ciertos órganos en particular, que puedan considerarse admisibles. 2) La evaluación de las cantidades absorbidas y de las cargas corporales de todos los núclidos pertinentes que podrían dar lugar a que la intensidad de dosis alcance o conserve los valores máximos'precipitados. 3) La vigilancia radiológica de los individuos expuestos, a fin de determinar cuál es la fracción de la carga corporal admisible de cualquier núclido retenida en el cuerpo. Las técnicas de vigilancia radio­ lógica basadas en la antropogammameuía o en la obtención de datos relativos a la velocidad de excreción • o'lala velocidad de exhalación, son susceptibles de tratarse en el simposio. • Entre los principales problemas planteados, figuran los siguientes: • a) Para la mayoría de los radionúclidos, no se dispone de datos metabólicos suficientes para relacionar las intensidades de dosis en los tejidos con las cargas corporales o las velocidades de excreción, o bien los datos disponibles no se basan en un número suficientemente elevado de sujetos humanos normales. b) Las grandes variaciones que se observan entre los diferentes sujetos en lo que respecta a funciones fisiológi­ cas tales como la velocidad de excreción o la velocidad de depuración, limitan seriamente la posibilidad de formular deducciones partiendo de la observación de casos aislados y restringe también la posibilidad de hacer deducciones relativas a la carga corporal partiendo de las velocidades de excreción observadas en un individuo determinado. ■ * c) La compleja ley de variación de la velocidad de excreción en función del tiempo, después de una absorción única de varios núclidos, impide deducir directamente la carga corporal de la velocidad de excreción cuando no se conoce el curso que sigue la absorción en función del tiempo, salvo en el caso especial de un núclido que se excreta con una velocidad que disminuye según una ley exponencial simple en función del tiempo, a contar del momento de la absorción. . . d) En particular, en el caso de muchos radionúclidos que revisten gran importancia porque se concentran considerablemente y quedan retenidos largo tiempo en ciertos órganos, se puede observar una rápida excreción inicial de la fracción absorbida que no se concentra de esa manera y una excreción mucho más lenta y prolongada de la parte retenida. De ese modo, una pequeña absorción reciente contribuirá más a la excreciónque una carga mucho mayor e importante que se haya localizado previamente en el órgano critico. ‘ e) Las fórmulas de la CIPR, que generalmente se basan en modelos exponenciales simples, pueden servir, como aproximación o bien a falta de información metabólica más detallada, para estimar las dosis atribuibles a absorciones o cargas corporales particulares, pero quizás no se presten tanto para relacionar la velocidad de excreción con la carga corporal en cualquier individuo en el caso de núclidos que se mezclan en sistemas de compartimientos múltiples en e l‘cuerpo humano, o son excretados de los mismos, particularmente en el caso de las velocidades de excreción’observadas-después de una absorción reciente. Estos casos exigen un examen especial, que se efectuará teniendo en cuenta los datos relativos a las fases preliminares del ciclo de renovación. . . 6 E. E. POCHIN

The prevention of undue occupational exposure to internal radiation in­ volves a number of steps. 1. Decisions must be made as to the highest doses, or dose rates, that can be regarded as perm issible, both for the body as a whole when it is more or less uniformly irradiated, and for individual organs or tissues under conditions when one or several of these are selectively irradiated. These decisions must be kept constantly under review in the light of developing quantitative information as to the type and amount of hazard as­ sociated with radiation exposure, and the factors determining the sensitivity of the' various tissues to radiation, particularly as regards the possibilities of lëukaemogenesis and carcinogenesis. 2. Estimates must be made, for all relevant nuclides, of the intake into the body, and of the total body content of each nuclide, which would cause any such doses or dose rates to be reached or sustained. These estimates also require continuous review, since they must often at present be based on quite inadequate information as to the distribution and speed of metabolism of many elements, at least as established in ade­ quately large numbers of normal human subjects. Moreover, the effects of differences in dose rate, of the specific localization of nuclides within organs, and of any gross non-homogeneity of tissue distribution are im ­ perfectly understood, and the relative biological effectiveness of different radiations for human carcinogenesis can usually be estimated only by in­ ference, very indirectly, from the production of other effects in other s p e cies. 3. Methods must be developed for monitoring exposed individuals, as well as their environment, to determine what fraction of a permissible body burden of any nuclide may be retained in the body. Papers in the present symposium deal particularly with techniques of monitoring by whole-body counting, and by evidence obtained from excretion or exhalation rates. I think that the holding of this symposium is particularly opportune at the present time and w ill be of value to the work of ICRP and to its task group concerned with this subject. I would like to express the Commissions' ap­ preciation to IAEA, ILOand WHO for arranging this symposium. Before discussing the relationship between the whole body or organ burden of various nuclides and the associated excretion rates, it is important to emphasize two general points. Firstly, we aré concerned with the practical problems of rapid and routine detection of contamination of individuals,, or of their more detailed examination if they are substantially contaminated. For many gamma- emitting nuclides, whole-body counting may be the most reliable, although not necessarily the most convenient, form of monitoring for body burdens approaching permissible levels, and we-are more concerned with a number of particular problems of excretion, especially involving alpha and pure beta emitters, than with the general question of excretory relationships for all elements. Secondly, although the observation of regulations implies a precision in procedure, the large variability between different normal subjects in physiological functions such as excretion or clearance rates, would severely limit the deductions as to body burden that could be drawn from excretion rates observed in any individual, even if the contamination were known to be TISSUE EXPOSURE AND EXCRETION RATE 7 due to a single intake- of the nuclide in known form on a known date, and if the metabolism of the elerriènt were known in detail for human subjects. Thus for example, various indices of renal function have been widely studied in human subjects, and show an individual variability, depending substantially upon body size. Even when expressed, however, per square meter of body surface, as estimated from height and weight, the variation is still substantial, four such conventional indices showing standard devi­ ations for individuals of from 17 to 25% of the mean values for all subjects (Table I) [1]. The variation has some dependence on age and is hère ex­ pressed in terms of mean values for an occupational age range of from 2 0 to 60. However, even if values are related to age by decades, the standard deviations still range from about 15 to 22% of the means corrected for age and body size, for the various renal clearance rates. Similar normal variability is observed for various measures of dis­ tribution space, standard deviations being in the region of 15% of the cor­ responding mean values, and for certain indices of respiratory function, with' rather higher variability, even though all such measures are referred to body weight or surface area, and are determined under carefully stan­ dardized conditions (Table II). Similar or greater variability in faecal ex­ cretion is to be expected (Table III) although any necessary measurement of daily faecal concentration of nuclides might, with no greater difficulty, in­ clude that of daily faecal mass, and eliminate much of the relevant varia­ bility. ' ■ 1 ■

TABLE 1 .

COEFFICIENTS OF VARIATION FOR RENAL CLEARANCES

Standard deviation as percentage of ¡mean value per mz body area

t A ll ages 20j t o -60 ■ By decades, 20 ю 60

Inulin clearance i . 15 1 Diodrast clearance 1 9 ! ■ 17

Para Amino Dippurate clearance 25 ' ' 22* .

U rea cle a ra n c e .„ ' 121 ’ 2 1 .

i

Such variability in-normal excretory rates will presumably apply to most elements, although some differences may be expected according to whether the element is or is not one which lis homeostatically controlled at constant concentration in the blood, or is chemically confounded with any such element. Little information.is yet aváilable on the normal variability in excretion rates for most nuclides in man, although urinary data for I 131, in effectively healthy subjects examined after thyroid ablation, show vari­ ation in the half period of radioiodine excretion which is comparable to that E. E. POCHIN

TABLE II

COEFFICIENT OF VARIATION FOR.DISTRIBUTION SPACES AND-PARAMETERS OF RESPIRATORY FUNCTION

. siVMa . ' ' e f t) Distribution spaces

Plasma volume (per kg) ' 13

Blood volume (per kg) 14

Thiocyanate space (per m2) . 16

Antipyiine space (per m2) 17

Respiratory

. Resting tidal volume 23

. Vital capacity ■ 22

Total lung volume 21

, Residual lung volume 17

15 s breathing capacity 31

a Standard deviation as a percentage of the mean. of the more conventional measures of renal function discussed above (Table.IV) [2]. These levels of variability clearly imply that the first measurement that we obtain in a human subject for the clearance or turnover rate for a nuclide may well be half or twice the estimate obtained from a second sub­ ject; and it is apparent that valid estimates of mean caesium turnover or radium excretion only emerged when appreciable numbers of subjects had been studied. It is also clear that the tissue dose commitment for an individual cannot be assessed with any accuracy from excretion studies, although with increasing knowledge it may be practicable to establish ap­ propriate and representative mean values for excretion rates corresponding to certain levels of body burden under defined conditions of intake, in the same way that the levels of body burden or ingestion of nuclides specified by ICRP are intended to correspond with certain tissue dose rates, at normal mean levels of metabolic activity, rather than at the extreme high or low levels of these activities which may be encountered at the limits of normal variation. . The step from tissue concentrations giving permissible dose rates, or even from the body or organ burdens taken as corresponding to these, to any associated mean rates of excretion, involves several obvious difficulties. The lack of adequate human metabolic data is as great in this field as it is in regard to the estimation of appropriate levels of body burden or of intake for many nuclides. .Even greater difficulties arise, however, from the com- TISSUE EXPOSURE AND EXCRETION RATE 9

TABLE III

DAILY FAECAL MASS D ry weight

(g /d ) No. of subjects

0 - 6

1 0 - 20

2 0 - 15

3 0 - . 16

4 0 - 13

5 0 - 9

6 0 - 6

7 0 - 5

8 0 - 0

9 0 -1 0 0 2 . .

1 6 0 -1 7 0 1

T o tal 93

Mean 37 g/d; SD..25 g/d; SD/M = 67^,

Note; subjects with "abnormal" fat excretion • of over 6 g/d were excluded.

plex way in which the excretion rate varies with time since intake for most elements which are absorbed into the body, from the gut or from the lung. Except when such an absorbed element is distributed through body tissues between all of which its equilibration is rapid compared with its rate of ex­ cretion from the body, the excretion rate is unlikely to decrease in a simple exponential way with time after a single intake, and the relationship between the excretion rate and the body burden w ill thus depend upon the tim e since a single intake, or on the recency and relative magnitudes of multiple in­ takes. For many important radionuclides, the retention of a small propor­ tion for .a long time in a particular tissue imposes the dosimetric limitations for protection purposes, but the rapid excretion of the majority of the in­ gested material from the rest of the body dominates the excretion ratesfol- lowing any recent intake. In all such cases, therefore,, excretion rates may be insensitive indicators of tissue dose commitment unless measured at some days or even weeks after any possible intake. This is clearly the case for calcium, strontium and probably many bone-seeking elements .with bio­ logical half periods of turnover in bone of several or many years but of a few days in the remainder of the body, or of iodine with half periods of a few months in the thyroid and of a few hours in the tissues otherwise. The urinary excretion of 1 цс of I 131 in one day could indicate a thyroid burden 10 E .E . POCHIN

TABLE IV

RADIOIODINE EXCRETION IN ATHYREOTIC SUBJECTS

Half periods No. of subjects (d)

0 .1 - 0

0 .2 - 1

0 .3 - 10

0 .4 - 33

0 .5 - 29

0 .6 - 16

0 .7 - ’ 2

0 .8 - 4

0 .9 - 0

1 .0 - 2 '

1 .1 - 0

.1 .2 - 1 .3 1

T o ta l 98

Mean 0.55 d; SD 0.16 d¡ SD/M = 2№ .

Note: SD/M would equal 24?» even if the three values greater than 0.9 were excluded as "abnormal”) . of i цс from an ingestion on the preceding day, or one of 2 0 0 /лс fro m an ingestion in the preceding week. The value of excretion measurements, par­ ticu larly in relation to any other crite ria that may be available, is thus very particular to the nuclide in question, and it is a useful aspect of the sym­ posium that various nuclides of importance are dealt with in detail. With regard to the general application to this problem of ICRP's re­ commendations and metabolic data, a number of points require discussion. ijor essentially insoluble materials, faecal examination will offer the only method of detecting the ingestion of radionuclides which give rise to no gamma radiation and which are not absorbed; and, at best, the method will detect only an ingestion during the preceding one or two days. The same remarks would apply for the inhalation of such materials, subject to the assumptions made as to the fractions of. inhaled insoluble materials which are coughed up and swallowed, 5/8 th of the total amount being assumed tobe r e ­ excreted in this way. These values are likely to vary with the particulate state of the inhaled material but offer a possible basis for detection of sub­ stantial exposure. Little is known as to the likely delays in the passage of particulate material down the gut, but the use of insoluble "m arkers ' 1 fo r TISSUE EXPOSURE AND EXCRETION RATE 11 determining transit time of material through the gut in metabolic investi­ gations indicates that such transit times are fa irly uniform and delays cannot be considerable. This type of monitoring would thus be mainly of any recent intake, although complicated by slower continuing excretion of fractions of initially retained material. For many nuclides ingested in soluble form, parts of the gut are re­ garded as the "critical" tissues in the sense that dividing cells in the gut wall are estimated to receive higher significant doses than other body tissues. For these nuclides, and even for many in which other body organs are regarded as critical from doses received from absorbed nuclides, the radioactivity of unabsorbed material in the faeces is usually very much greater than the radioactivity of material which is re-excreted by all routes after being absorbed into the body. Figure 1 shows the relationship, ' on metabolic models currently adopted by ICRP [3], between the activity which is excreted unabsorbed in the faeces, and that which is re-excreted after absorption, for various nuclides in soluble form for which the gut is regarded as critical, and at the maximum perm issible levels for continuous intake. Figure 2 gives comparable data on those for which other tissues are critical. It is evident that for almost all nuclides, and excepting eleven elements which are treated as being fully absorbed, detection of re-excreted

• Fig. l

Activities excreted unabsorbed, and re-excrcted after absorption, for soluble but incompletely absorbed nuclides having parts of the gut as first critical organ, when intakes are by ingestion at maximum permissible levels. "Unabsorbed" activities estimated from ICRP MPCW values (166 h-week)x2200 m l/d x(l-f,) unabsorbed.

Re-excreted activities based on maximum permissible body burdens for the organ (other than gut) which is most limiting, . the ratio of values of MPCW estimated for gut and for that organ, and the biological half-life applicable to the total body burden. 12 E. E. POCHIN material will involve higher, and often very greatly higher, sensitivity than the detection of unabsorbed material, even ignoring the ten-fold greater average daily bulk of urine than of faeces.

Fig. 2

Activities as for Fig. 1, but for nuclides for which gut is not the first critical organ.

Re-excreted activities based on the limiting body burden and the biological half-life for the total body burden.

The detection of re-excreted material, commonly in urine, will, how­ ever, often have substantial advantages apart from the awkwardness of arranging routine faecal assays, since the limits of detection are frequently adequate at well below permissible levels and particularly since, for any substantial retention times in the body, the examination will detect intakes occurring over'longer periods than only a day or two, and so need not be used so îr e r quently. The method is of course the only one available for nuclides which aré fully absorbed and without gamma emission (namely - on the Commission's criteria - Н З , СИ, F ie, S35, C136, Rb87 andCsiss). ■ ICRP models for the distribution and turnover of nuclides have been formulated in order to relate various levels of continuing intake or of body burden of these nuclides to maximum permissible dose rates or dose com­ mitments in the body as a whole or in particular tissues. There are a number of limitations to their use for the estimation of excretion rates which would correspond with particular dose rates in tissues, even when the time or pattern of intake of a nuclide was known. In particular, the short stay of a large fraction of an absorbed radionuclide in the body may make a small contribution to tissue doses as. compared with longer retention of smaller fractions, especially as averaged over the relevant integrating periods of three months or longer used by ICRP. This transient phase of the element's metabolism may therefore, appropriately, not be reflected TISSUE EXPOSURE AND EXCRETION RATE 13

in the highly simplified form of model that is adopted for dosimetric pur­ poses, even though it may make a predominant contribution to the pattern of excretion within a short period'after intake. The models are in general expressed in the form of that single exponential function which, in the light of available information, is believed best to express the partition and cir­ culation of nuclides through body tissues. Even for the purposes of dosi­ metry, it is clear that this represents a considerable over-simplification when one considers the known complexities, for elements of which the human metabolism has been closely studied, of varying absorption of the different chemical forms, plasma protein binding of compounds, multiple body com­ partments with differing rate constants of equilibration, re-excretion into the gut and re-absorption from it, passage of the element through metabolic cycles with differing organ concentrations for different forms, concentration in zones of an organ, having differing turnover rates, varying degrees of re­ utilization of a compound on its release from cells of an organ or from bone crystals, and translocation within the body of an element or its daughter products. The metabolic models for all elements, apart perhaps for those with the most general distribution and the most rapid equilibration through­ out body tissues, may require extensive revision as more detailed human metabolic data become available or closer approximations become useful. Even these revisions, however, will be primarily directed towards the re­ latively long term dose commitments of intakes or body contents of .various forms, and the interpretation of excretion rates appears to require formu­ lations of different type and approach. In this connection, emphasis should begiven, in the study of nüclides of practical importance for which good alternative determinations of body burden are not readily available, to the size and turnover of those components of intake which are rapidly excreted, with evidence as to the time by which their excretion is substantially com­ plete and the time, after a single intake, at which excretion or exhalation begins to reflect the body content of, those fractions of the intake which con­ tribute mainly to the limiting tissue doses. For such purposes the use of "power functions" may be considerably preferable to the use of the single exponential which most appropriately summarizes the long term dose com­ mitment to the critical organ (and even this will not necessarily be equivalent in time constant to that used for the whole body, and from which the overall loss from the body can be derived). It must bé emphasized however that the power function will in general approximate to a description only of excretion rates for which the size of the various components decreases pro­ gressively in proportion as the half periods applying to these components increase. If 50% of an absorbed nuclide is excreted with a half period of 1 d, 25% with a half period of 4 d, and 12^% with 16 d, a pow er function of form y = at-i.5 will describe the excretion rate rather accurately from a few days to a few weeks. This will however give no assurance that, for example, the next 6.25% will have a half period of 64 d'and a further 3.125% one of 256 d, so that the same power function will continue to be followed; and it may be these later, smaller, and slower components which correspond to the long retentions in the critical organ which determine the limiting dose. Neat mathematics is no substitute for adequate human metabolic data. I think we therefore require, and it will be a valuable function of this symposium, and of the task group of ICRP concerned with this subject, to 14 E. E. POCHIN develop information which not only relates the total body content of nuclides more securely to the maximum resultant dose rate in different tissues and at different times after absorption, but also defines the routes and rates of excretion of nuclides in relation to the time of their ingestion or inhalation in known amount and in determined chemical form. It may be valuable to know a number of factors: the ratio between urinary and faecal excretion rates, and the evidence that this ratio may give as to the recency of any major contamination; the chemical form of the excreted material, which could in some circumstances date the intake; the excretion rates associated with tissue burdens resulting from wounds; or the degree of approximation with which the excretion rate at a stated period after the last possible intake, or the relationship between rates at different such periods, can be taken to indicate the body or organ content of a nuclide or the dose commitment due'to them. The necessary information is sparse enough for many of the nuclides of practical importance, and for which no alternative assessments of body burden are available. For others of less frequent occurrence, even the simplest information on human excretory rates is lacking.

REFERENCES

[1] SPECTOR, W. S. (Ed.), Handbook of-biological data, Saunders, Philadelphia and London (1961). [2] POCHIN, E .E ., Proc. 8th Conf. of Italian Soc. for nucl. Biology and Medicine, Pisa, 1963, in press. [3] Recommendations of the International Commission on Radiological Protection, ICRP Publication 2. Report of Committee II on permissible dose for internal radiation (1959). Pergamon Press (1959); Health Physics 3 (June 1960).

DISCUSSION .

R.D. JORDAN: Would it not be more accurate, mathematically, to use a sum of exponentials to describe a complex excretion process which, in­ volves several mechanisms, rather than to use a simple power function? E.E. POCHIN: I would certainly agree that excretion rates are likely to be best analysed as a series of exponentials, when excretion is occurring from various body "compartments" with different turnover times. When, however, the size of these compartments is broadly related to their turnover times in the manner that I mentioned, a single power function may adequately fit the sum of these exponentials. In these circumstances, which frequently appear to hold for human data, a power function may conveniently summarize the excretion process, even though it is valueless for physiological inter­ pretation of the data and unreliable for any extrapolation beyond the range over which it has been validated. ON THE DEVELOPMENT OF DEVICES FOR THE DETERMINATION OF TOTAL-BÔDY RADIOACTIVITY IN MAN: A HISTORICAL AND CRITICAL REVIEW

B. RAJEWSKY, A. KAUL AND J. HEYDER MAX-PLANCK-INSTITUTE FOR.BIOPHYSICS, FRANKFURT/MAIN . FEDERAL REPUBLIC OF GERMANY

Abstract — Résumé — Аннотация — Resumen

ON THE DEVELOPMENT OF DEVICES FOR THE DETERMINATION OF TOTAL-BODY RADIOACTIVITY IN MAN: A HISTORICAL AND CRITICAL REVIEW. With the increasing application of artificial and natural radioisotopes, mainly in industry, scientific research and medicine during the past years, the in vivo deter­ mination of the total-body content of incorporated radioactive substances became important at a rather high degree. On the one hand,, it was the purpose of investigations to determine the natural content of radionuclides within the human body to obtain fundamental values for the fixation of "maximal body burden". On the other hand, it became necessary as a task of radiation protection to survey continuously the employees of, e .g . reactor stations, isotope laboratories and clinical hospitals, so that even the lowest incorporations could be detected as early as possible. . The determination of the amount of incorporated radionuclides by external direct measurements of the radioactivity goes back to the early 1930s, when the first cases of radium poisoning were examined by means of ionizacion chambers. Only a few years later a first "Institution for the Physical Diagnosis of Radium Poisoning” was established at the former Kaiser Wilhelm Institute for Biophysics in Frankfurt (M ain). This institution carried out in vivo and post mortem diagnoses. In vivo diagnosis consisted mainly of determination of the excretion of Ra226 and its daughters in faeces, urine and the breath. In addition, distribution of the incorporated Ra226 within the body was examined by means of ionization chambers. , However, one of the problems of radiation protection, namely recognition of even low radioactive incorporations as early as possible, could not be taken into account at that time, since the instrumental limit of detection was of the same order of magnitude as the maximum permissible body burden for Ra2*6 (Q. i ^g/total body). However, it became possible about 15 years ago to determine radioactive incorporations of about one order of magnitude only below the normal Ra226 body content by means of high-pressure ionization chambers and, later on, by scintillation counters. Furthermore, mainly during recent years, the, application of methods for the external direct examination of incorporated radionuclides was extended to clinical and radiological problems concerning the metabolism of different radionuclides. The experimental possibilities of the direct and indirect determination of radioactive incorporation within the human body are discussed in this paper, which is an historical and critical review based on the authors’ investigations and those of others. -

CONSTRUCTION DE DISPOSITIFS PERMETTANT DE DETERMINER LA RADIOACTIVITÉ TOTALE DE L’ORGANISME HUMAIN: APERÇU HISTORIQUE ET CRITIQUE. Les applications des radioisotopes naturels et artificiels n’ayant cessé de se multiplier au cours de ces dernières années, notamment dans l'industrie, la recherche scientifique et la médecine, il est devenu très important de mesurer in vivo la charge corporelle totale de substances radioactives incorporées. D’une part, l'objet de certaines recherches a été d’établir la charge corporelle normale de radionucléides on vue d'obtenir des valeurs fondamentales pour fixer la «charge corporelle m aximum » , D'autre part, aux fins de radioprotection, il a fallu commencer à exercer une surveillance continue des travailleurs des centres nucléaires, laboratoires de radioisotopes, hôpitaux, e tc ., afin de pouvoir dépister le plus rapidement possible les moindres quantités incorporées. La détermination de la charge corporelle d'un radionucléide par des mesures externes et directes de la radioactivité remonte au début de la décennie 1930-1940; à cette époque, on a utilisé des chambres d’ionisa­ tion pour examiner les premiers cas d'empoisonnement par le radium. Quelques années plus tard seulement, une première «institution pour le diagnostic physique de l’empoisonnement par le radium » a été créée auprès

15 16 В. RAJEWSKY et al.

de Tancien Institut Kaiser Wilhelm de biophysique, à Francfort-sur-le-M ain. Cette institution s’occupait à la fois de diagnostics in vivo et d’autopsies. Le diagnostic in vivo consistait essentiellement à déterminer l’excrétion de 226Ra et de ses produits de filiation dans les matières fécales, les urines et l'haleine. On étudiait en outre la répartition dans l'organis­ me du 226Ra fixé, au moyen de chambres d’ionisation. Cependant, à cette époque il n'était pas possible de s’occuper de l’un des problèmes de la radioprotection, savoir celui qui consiste à déceler le plus tôt possible la présence de quantités même faibles de substances radioactives incorporées; en effet, la limite de détection des appareils était du même ordre de grandeur que la charge corporelle maximum admissible de 226Ra (0,1 ^g/organisme entier). Il y a 15 ans environ, il est devenu possible de déceler des quantités incorporées inférieures d’un ordre de grandeur à la charge corporelle normale de 226Ra; pour cela, on s’est servi de chambres d’ionisation sous haute pression et, plus tard, de compteurs à scintillation, • Durant ces dernières années, on a également appliqué des méthodes qui consistent à mesurer directement .du dehors les quantités incorporées, pour étudier les problèmes radiologiques et cliniques que pose le métabo­ lisme de plusieürs radionucléides. Les possibilités expérimentales de mesurer directement et indirectement des substances radioactives fixées dans le corps humain font l’objet d'un aperçu historique et critique fondé-sur les travaux de l'auteur et d’autres chercheurs.

О РАЗРАБОТКЕ ПРИБОРОВ ДЛЯ ОПРЕДЕЛЕНИЯ РАДИОАКТИВНОСТИ ВСЕГО ОР­ ГАНИЗМА У ЧЕЛОВЕКА ; ( ИСТОРИЧЕСКИЙ И КРИТИЧЕСКИЙ ОБЗОР). Ввиду возросшего использования в последние годы искусственных и естественных радиоизотопов в промышлен­ ности, научных исследованиях и медицине определение in vivo общего содержания инкорпори­ рованных радиоактивных веществ в организме приобрело особо важное значение. С одной, стороны^это было предметом исследования естественного содержания радиоизотопов в чело­ веческом организме для получения фундаментальных данных относительно фиксации "макси­ мального содержания в организме". С другой,— это стало необходимостью, частью системы радиационной защиты для постоянного наблюдения за служащими атомных электростанций, изотопных лабораторий и клинических госпиталей, для наиболее раннего обнаружения даже очень малых количеств инкорпорированных изотопов. Определение количеств инкорпорированных радиоизотопов с помощью прямого наружного измерения уровня радиоактивности началось в 1930 г ., когда исследовались первые случаи отравления радием с помощью ионизационной камеры. Несколько лет спустя в бывшем Инсти­ туте биофизики им. кайзера Вильгельма во Франкфурте на Майне было создано первое "уч­ реждение для физического диагноза радиевых отравлений" . В задачи этого учреждения вхо­ дила диагностика как in. vivo, так и post mortem. . 1 Диагностика in vivo включала, главным образом, определение выделения Ra226 и про­ дуктов его распада в экскрементах, моче и выдыхаемом воздухе. Изучалось также распре­ деление инкорпорированного Иа22й в организме с помощью ионизационной камеры. Однако одна из проблем радиационной зашиты — определение малых количеств инкорпори­ рованных в наиболее ранние сроки не могла быть решена в это время, поскольку пределы об­ наружения для приборов совпадали с величиной максимально допустимого содержания Ra226 в организме (0,1 мг во всем организме). Но 15 лет назад оказалось возможным определение содержания инкорпорированных радиоизотопов лишь на один порядок величины ниже нормаль­ ного содержания Ra226 в организме с помощью ионизационной камеры высокого давления, а позднее — сцинтилляционных счетчиков. Кроме-того, главным образом в последние годы, методы прямого наружного измерения инкорпорированных .радиоизотопов были использованы для решения клинических и радиоло- гиееских проблем, касающихся обмена различных радионуклидов. Обсуждаются экспериментальные возможности прямого и непрямого определения ин­ корпорированных радиоизотопов в челеовеческом организме, основанные.на литературных данных и данных автора.

PERFECCIONAMIENTO DE INSTRUMENTOS PARA DETERMINAR LA RADIACTIVIDAD TOTAL DEL ORGANISMO HUMANO: ESTUDIO HISTÓRICO Y CRÍTICO. Debido a la creciente aplicación de los radio­ isótopos naturales y artificiales, principalmente en la industria, investigación científica y medicina, ha aumentado considerablemente en importancia la determinación in vivo del contenido total en el cuerpo de sustancias radiactivas a él incorporadas. El propósito de las investigaciones fue, por una parte, determinar el contenido natural de radionúclidos en el cuerpo humano para obtener los valores fundamentales en que DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 17

basar la fijación de la fijación de la «carga corporal m áxim a». Por otra, la necesidad, en el marco de la protección radiológica, de examinar continuamente a los empleados 'de reactores nucleares, laboratorios de isótopos y hospitales clínicos, entre otros, para detectar lo más precozmente posible la más mínima in­ corporación de radionúclidos. La determinación de radionúclidos incorporados por medidas externas directas de la radiactividad se remonta a principios del decenio 1930-1940, cuando se examinaron los primeros casos de envenenamiento por radío con cámaras de ionización. Sólo unos pocos años después fue establecido el primer «instituto para el diagnóstico físico del envenenamiento por radio » , en el antiguo Instituto Kaiser Wilhelm de Biofísica en Francfurt (Main), instituto que se ocupaba no sólo del diagnóstico in vivo, sino también del diagnóstico post mortem. , . El diagnóstico in vivo se refería preferentemente a la determinación de. la excreción d e2z6Ra y descen­ dientes por las heces, orina y aire exhalado, y a estudiar su distribución en el organismo por medio de cámaras de ionización. En aquel tiempo, sin embargo, uno de los problemas de là protección radiológica, a saber, él reconoci­ miento precoz de mínimas cantidades de sustancia incorporada, no pudo encararse porque.el limite instrumental de detección era del mismo orden de magnitud que la carga corporal máxima admisible, para el 226Ra (0,1 |jg en todo el cuerpo); pero, aproximadamente hace 15 años, ya se comenzó a efectuar la determinación de sustancias radiactivas incorporadas tan sólo de un orden de magnitud por debajo del contenido normal de 226Ra, utilizando cámaras de ionización de alta presión y, más tarde, contadores de centelleo. En los últimos arlos, además, la aplicación de métodos de examen directo externo para detectar radio- núclidos incorporados se ha hecho extensiva a problemas clínicos y radiológicos referentes al metabolismo de diferentes radionúclidos. • , Las posibilidades experimentales de la determinación directa e indirecta de incorporaciones radiactivas en el cuerpo humano se examina en ese estudio histórico y critico basado tanto en trabajos de investigación del autor como en trabajos de otros especialistas. ’

INTRODUCTION . .

Radioactivity was discovered 65 yr ago, the first signs of radium poison­ ing after the incorporation of radioactive material being observed nearly 25 yr later. At first, only some disquieting communications existed. These communications occurred more frequently and resulted in the scientific problem relating to body damage after the incorporation of radioactive material being increased. Today this problem is very acute, a special characteristic being the high toxicity of the radioactive m aterial. This toxicity is not the normal toxicity of pharmacy but the toxicity which results from ionizing radiation produced in the organism. First investigations of radium poisoning had already demonstrated that very small amounts of radioactive material de­ posited in the body lead to serious damage of the organism and later on to death. It also became apparent that the problem of radium poisoning is very complex and, even today, a fter working 30 y r upon it, many questions are still not resolved. It is interesting to note that when radioactivity was first discovered it was regarded as a curing and animating property fo r human beings,. From this arose the "Radium-Schwachtherapie" (1905-1908) and the "Radium- Balneologie". The curing efficiency of the water of many springs in -which traces, or small amounts of radioactivity were found was thought to be due to the effects of the radioactive substances. The first dramatic case of ra­ dium poisoning was caused by the drinking of radioactive water. The case was given considerable attention and led to the relevant research. A very wealthy American citizen, John Bird, believed in the curative properties 18 В. RAJEWSKY et al.

of radioactive water and drank this water only, with the result that he died some years later from acute damage from the effects of the water. Not­ withstanding this, many famous health resorts, based on radioactive therapy, still exist today and thousands of people find alleviation or cure there. Ob­ viously, Mr. Bird had far exceeded the tolerance limit of the daily intake of radioactive material. Luminous dial painters in the watch industry also showed symptoms of radium damage, as did patients who had been injected, for experimental purposes, with the same or greater amounts of radioactive solutions as Bird. These cases induced research on and scientific analysis of radium poisoning in the United States in 1932/33. At the same period three lethal cases of radium poisoning were established in the radium manufacturing industry in Germany which were investigated by the pathologists Fischer- Wasels and collaborators and by the biophysicists Rajewsky and collaborators. After the preliminary studies had been undertaken with Geiger tubes and ionization chambers, Rajewsky proposed the following: (a) the development of highly sensitive devices for the determination of radium deposits in the human body; and (b) determination of the. lowest toxic amounts of radium in the hùman body anci, establishment of the maximum permissible level for radium-226. Since then investigations carried out at the Max-Planck Institute for Biophysics in Frankfurt have been devoted to these problems. First clinical investigations indicated a very complex syndrome of radi­ ation disease or damage,- the most important symptoms being malignant neoplasm, fibrosis of the lungs and leukaemia. These symptoms, including induction of cancer after the application of external irradiation in therapy, were already known at that time. From this Rajewsky connected his own investigations on radiation cancer with radium poisoning, and decided to undertake the experimental production of cancer caused by small amounts of incorporated radioactive material. For more than 300 yr there was the problem of the "Schneeberger" and "St. Joachimsthaler Bergkrankheit" (mainly cancer of the lungs). Investigations also covered this problem. Chief interest was focussed on the radionuclides radium-226, radon-222 and MsTh1 (radium-228), which were then used in industries producing radioactive pre­ parations, luminous paints, etc. Earlier the "Schneeberger-Bergkrankheit" was already connected with the inhalation of radon, but this was a supposition only. Since the beginning of the atomic age radioactive substances are being produced and used in large amounts which has appreciably altered health hazard. Before the use of artificially-produced radioactivity a small percentage only of the population was in danger of incorporating radioactive materials but today, as a result not only of the development and, testing of atomic bombs, but also of the extensive peaceful uses of radiation in science and industry, a large proportion is liable to incorporate such substances involuntarily. Before the discovery of nuclear fission very few research teams in­ vestigated the pertinent problems, but today in most countries large groups of scientists are involved. Within the last decade progress has been con­ siderable, particularly in the technique of measurement, but our knowledge DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 19 in regard to the radiobiological, medical and clinical aspects is not complete, among the questions still unresolved being that of knowledge of the lowest dangerous amounts of radioactive material and the effect of fractionation of_the dose. At present, the situation is as follows: (1) By modern techniques of measurement practically every radioactive deposit in the human body, no matter how small, can be measured in vivo or post mortem, although the technique of localizing such deposit is not yet fully efficient. Quantitative distribution of the incorporated substances is only shown by biopsy or post-mortem investigation. (2) One of the most important questions is that of the nature, physico­ chemical property and size of the carrier of the radioactive substance before incorporation. * (3) The biophysical and physiological mechanisms of the transport of radioactivity and . its action within the body, as well as dose and clinical symptoms of internal inhomogeneous irradiation are not fully, or, even partially, understood. . (4) There is thus no absolute prQof of the amount of permissible dose or permissible concentration in the body. The presently-accepted values are based on reasonable extrapolated data of experience, but should in future be on a better experimental.basis. . . . . ■ The problem mentioned above under (4) is the most important and the need of development towards higher precision of the technique of determi­ nation of radioactive substances depends on its solution, as also do the prob­ lems mentioned under (1), (2) and (3) above. The first guiding determination of the lowest dangerous amount of de­ posited radioactive substances was obtained by simpler experimental methods than those in use today. In 1934 FLIÑN [1] presumed deposits of 1 дс radium-226 to cause damage at the place of deposit. In 1933/34 RAJEWSKY [2, 3] arrived experim entally at the still toxic amount of 0.75 juc radium-226. He thus postulated the amount of radium-226 which could be incorporated without danger to be less'thán 1 ¡ug ("Restradium"). Today, 30 yr later, this value is still valid. In 1941 the value of 0.1 y.c radium-226 was laid down to be the maximum permissible body burden [4] and was confirmed to bé valid at the International Congress of Radiology in London in 1950. This value was at the same time catalogued in the Recommendations of the International Commission on Radiological Protection [5]. It is still con­ sidered to be valid today and the maximum perm issible amount of long-lived alpha-ray emitters which are bone seekers, like radium, has also recently been catalogued (International Congress of Radiology at Copenhagen in 1953). The maximum permissible amount of other radioisotopes is estimated from the biological data known at present. , . These data are composed of the method of intake of the isotope into the body, the distribution pattern, the deposition and excretion of the isotopes and the results of metabolic research in consideration of the. physical data of the radioisotopes (kind of radiation emission, its energy, the portion of ab­ sorption and the physical half-life). Here the distribution pattern, the ex­ cretion and deposition in the body is assumed to be equal fo r both the radio­ isotope and the corresponding stable element. Furthermore, the permissible weekly absorbed dose for the life period is presumed to be constant and con­ tinuous and the body burden of the critical organ never to exceed a value 20 В. RAJEWSKY et a l.

as high as 0.3 rem. The performance of these estimations is not discussed here as they are covered more fully elsewhere [ 6 ]. Establishment of these fundamental values was 'the basis of the work carried on by the scientific group at Frankfurt (the Kaiser-Wilhelm-Institut für Biophysik) who developed sensitive detection methods and devices for radiation up to the recognized maximum permissible limits (1936/38). At the completion of this phase investigations were extended to the Schneeberger lung cancer, measurements being made and statistical and medical in­ formation being obtained on involutions in the mines and in laboratory ex­ periments with animals. A method was also developed for the physical and medical surpervision of those exposed to hazard. In 1938 the "Untersu- chungsstelle für die physikalische Diagnostik der Radiumvergiftungen" was founded at the Kaiser-Wilhelm-Institut for Biophysics in Frankfurt/Main. The diagnostic method for determination of radium poisoning included the following: ' (a)i- measurement of the gamma-rays emitted from the whole body and those emitted from specific'areas of the body; (b) measurement of radioactivity of excreta, respiratory air, urine and fa eces; • - (c) investigation of the blood, namely, blood count, sedimentation of corpuscles and the radioactivity of the blood; • (d) determination of the frequency and volume of breath; and (e) diagnosis by X-rays and clinical investigations, where necessary. From 1938 to 1944,the laboratory of the Untersüchungsstelle für die Physikalische Diagnostik der Radiumvergiftungen-carried out investigations on workers in the radium manufacturing industry and luminous dial painters in the watch industry. Where radium poisoning occurred the workers were permanently controlled. Research, of the Schneeberger Bergkrankheit and the aforementioned Untersuchungsstelle indicated causal connection between the radioactive deposits and formation of cancer. Efficiency of small amounts of deposited radioactive material was also indicated. , About the, same time researchers in the United States, especially R. D. EVANS [9], were developing highly sensitive’detection methods for radioactive deposits. The best results were obtained by Evans with his Geiger- counter device. In considering the radiation burden of the human body due to radioactive deposits, three factors are important, namely, ■ (1) Time of the introduction of radioactive material into the body and its course in time in tissues and organs. ■ (2) The course in time of the excretion of the radioactive material from the body/ ■■ (3) Increase in the enrichment of radioactive substances in different . , tissues and parts of the body in time, particularly in the critical organs. ' ■ Tothis should, of course, be added knowledge of the kind of radioactive substance (element, chemical compounds, etc.), particulars of the" indi­ vidual, his age, stage of development, state of health, life-long habits and profession, and so on. DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 21

1. DISCUSSION OF THE LOWEFl LIMIT WHEN MEASURING INCORPORATED RADIONUCLIDES SUCH AS RADIUM-226: A HISTORICAL REVIEW ' '•

1. 1. The indirect method. Measurement of excreted radioelements

The radionuclide in the body circulation is gradually excreted by con­ tinuing exchange with stable isotopes of the same or diadochol elements. The incorporated activity may be. estimated from the daily amount being excreted by faeces or urine and from the effective half-life. The measuring process, however, is subject to a main difficulty, namely, the amount of excreted radionuclides with a long effective half-life is very low. Only about 0 .0 1 % of the radium incorporated a few months before being fixed in the body as the so-called residual radium will be daily excreted with faeces and urine [ 8 , 9, 10, 11] . This means, in other words, that if one-tenth of the maximum permissible body burden of radium-226 is deposited in the skeleton only about 1 0 -1 2 с are expected to be found in the daily excrements. Such very small activities may be measured only with high expenditure of time and measuring devices. ’ • . The incorporated radium may also be measured by its exhaled decay product, radon. The radon concentration in the expired air was first measured in Germany fry a device which was developed by JANITZKY in 1933 for radium balneological purposes [12] . This device, which was modified for measurements of very small amounts of emanation [7] is fundamentally composed of two parts, namely, an apparatus in which the exhaled a ir is gathered and processed, which consists of a spirometer of 1 0 - 1 capacity, and a drying device; and . a measuring apparatus (see Fig. 1) consisting of an ionization chamber (l.'9-l capacity) switched in series with an electrometer. This device allowed of measurement of concentrations as low as 1 X 10-12 c radon pér litre. Assuming that only 10% of the radon built up in the human body is exhaled - at a respiratory rate of 10 1/m.in [7] - the above­ mentioned sensitivity is able to detect and determine any radium deposit in the human body as low as 0.5 pic. This low amount was, however, not sufficient for early recognition of incorporation. For this reason a new method was tried in order to lower the detection limit to the order of magni­ tude of the amount of radon corresponding to the maximum permissible amount of incorporated radium. This was attained by the use of an electro­ meter arrangement of special precision similar to HALLADENER [13]. Two ionization chambers of 2 - 1 capacity were switched in compensation to reduce the background caused by cosmic and environmental radiation [14]. With this device (see Fig. 2) it became possible to detect radon concentrations as low as 1X10_lo сfl­ it the above-mentioned rates of exchange and exhalation are assumed, this valye is equivalent to 0.05 /лс of radium-226 in thé whole body. If 10-1 cham­ bers are used, the limit of detection is lower by a factor of two. A control measurement was performed with the use of an alpha pulse ionization cham­ ber similar to that published by HURSH and GATES [15] having a limit of detection in the same order of magnitude [16] . The device worked without gas amplification, using the same chamber as the double ionization chamber in compensation. В. RAJEWSKY et al.

Ionization chamber device for the measurement oí concentrations up to about 1 x 10"12 c/1 [ 12]

F ig. 2

• 222 Ionization chamber device in compensation for the measurement of Rn concentrations up to about lx 10"13 c/1 [14] DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 23

In the field of scintillation counting there has been good progress as to the quality of photomultiplier tùbes and the purity of the scintillation material. Therefore there were good reasons to develop scintillation coun­ ters for the detection of very small radon concentrations. With reference to similar counters described bÿ MEGY et al. [17] and BYRANT et al. [18] it became possible to detect radon concentrations as low as 7 X 10-14 c/1 by means of a spherical scintillation counter (Fig. 3(a)) [19]. The measuring tim e was 4 h and the m aximum e r r o r ± 20%. At another c y lin d rica l scin­ tillation chamber (see Fig. 3(b)) only the window was coated with silver- activated zinc sulphide. The scintillation screen was vacuum-evaporated with a 0 . 2 - 0.5 ^m-thick aluminium layer from which the positive charged daughters were separated [19] . Such scintillation counters are more sensi­ tive than the ionization and pulse-ionization chambers. Besides this they have the following advantages of being proof against shocks, and the back­ ground is not elevated by environmental 7 -radiation.

Fig. 3a

Spherical ot-scintillation counter for the measurement of Rn222 concentrations up to about 7 X 10 ”14 c/1 [19]

Although the sensitivity is as low as 7X 10-14 с radon per litre, a radium content being greater than IX . 1 0 '8 с may be detected with the assumptions made as detailed above. If a routine control is concerned, however, with persons handling radium professionally, the exhaled radon needs to be en­ riched prior to the measurement, if its concentration is smaller than IX 10_1 3 c/l. That is why a device was developed [19] which is similar to that used in the United States [20] . The enrichment is performed by ad­ sorbing radon by previously-cooled charcoal. The lower limit of detecting radon is 5X 10-15 c/1 (corresponding to a whole-body radium-226 content as low as IX 10-9 c) caused by the naturally-occurring radium in drying ma- 24 В. RÀJEWSKY et al.

Fig. 3b

Cylindrical a-scintillation counter with electrostatic separation of Rn222 daughters . (Limit of detection: 7 x l0 "14c/l) [19] terial, the carbon dioxide adsorbents, the charcoal and the pressed aged tank air. . In Table I the measuring methods are summarized covering the de­ termination of the total-body content of radium-226 via the radon concentration in exhaled air. The summing-up is done according to historical sequence. In Table II the lower detection limits are summarized in a similar way. Herewith the detectable whole-body content of radium-226 is estimated on the assumption of a 1 0 % exhalation of radon and a breathing rate of 1 0 1/min. Up to now, in estimating the still detectable total-body content of radium-226 being obtained by measuring the exhaled radon, an emanation-exhaling coef­ ficient of 10% has been assumed [7] . According to experience this per­ centage is subject to great individual variations due to the time the radium remains in the human body, and to some physiological factors, during the time of collection of air [34] . Nevertheless in the case of a long-period deposition of radium, an average excretion of 70% [30] is assumed, a value being laid down in the Report of the United Nations Scientific Committee on the Effects of Atomic Radiation.

1. 2. The direct method. Measuring the gamma-rays or bremsstrahlung emitted by the human body

The principal disadvantage of the radon method is the uncertainty of the magnitude of the emanation exhalation coefficient, as determination of the radium deposits is based on the assumption of a determinate degree of radon. The measurement of the concentration of radon in the exhaled air merely DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 25

TABLE I

DEVICES FOR THE MEASUREMENT OF Rn 220 BY SEVERAL AUTHORS IN HISTORICAL SEQUENCE

■ Author (yr) D evice

SCHLUNDT, H. et al. [21] 1933 Ionization chamber

EVANS, R.D. [22] 19 3 3 Ionization chamber

EVANS, R.D. [23] 1935 Ionization chamber

RAJEWSKY', B ., JANITZKY, A. [7,12]s 1 9 3 3 /3 6 Ionization chamber

EVANS, R.D. [9] 1937 Ionization chamber

HOECKER, F.E. [24]. 19 4 4 - - .

LANG, K. et al. [25] 1947 - -

HARLEY, J.H. et al. [26] 1951 Ionization chamber

. MARINELLI, L.D. et al. [27] 1953 Adsorption of radon on activated charcoal ionization chamber

HURSH, J.B. [28] ' 1954 Adsorption of radon on activated charcoal a-ray pulse - ionization chamber

BATE, G.L. et al. [29] 1954 Ionization chamber

STEHNEY, A. F. et al. [20] 1955 Adsorption of radon on activated charcoal ionization chambers, proportional counter

NORRIS, W. et al. [30] 1955 - - '

MUTH, H. et al. [14] 1956 Ionization chamber in compensation

HANTKE, H.J. [16] 1958 a-ray pulse ionization chamber

HARLEY, J.H. et al. [31] 1 958 - -

MEGY, J. et al. [17]’ : 1958 Spherical a-ray scintillation counter =• adsorption of radon on activated charcoal from 100 1 expired air

KAUL, A. et al. [19] 1959 Spherical a-ray scintillation counter; a-ray scintillation counter with electrostatic separation of radon daughters: a-ray scintillation counter and adsorption of radon from 100 1 expired air

ROSS, D.A. [32] 1960 a-ray scintillation counter

HURSH, J.B. [33] 1963 Adsorption of radon on activated charcoal (continuously) and a-ray scintillation counter

indicates the minimum amount of incorporated radioactive material. Only t-he simultaneous measurement of the exhaled radon and of the gamma-ray radiation emitted from the human body makes it possible to give a statement on the total radium deposit. Besides this, the gamma-ray method has an TABLE II

DEVELOPMENT OF DEVICES FOR TIIE MEASUREMENT OF EXPIRED Rn 220 SINC E 1933 AND ESTIMATION OF THE DETECTADLE Ra226 TOTAL-BODY CONTENT

Limit of detection Total-body Ra226 content A uthor (yr) D ev ice (¡1C lin/1) (Me)

R A JFW S K Y . Í3. , 1 9 3 3 / 3 6 Ionization chamber 1 X 1 0 ' C - 0 . 5 JANITZKY, A. [7.12]

BATE, G.L. et al. [29] 1 9 5 4 . Ionization chamber 2. 5 X 10'9 • ~ 0 . 0 0 0 5 RJWK e al. et RAJEWSKY.

MUTH, II. et al. [14] 19 5 6 Ionization chamber in compensation 1 X 10 ~1 ~ 0 . 05

HANTKE, H.J. [IB] 1 9 5 8 a-ray pulse ionization chamber 1 X 1 0 ' 7 - 0 . 0 5

. MEGY, J. et al. [17] 1 9 5 8 S p h erical a-ray scintillation co u n te r 1 X 1 0 ' 8 ~ 0 . 0 0 5 adsorption of radon from 100 1 expired air

KAUL, A. et al. [19] 1 9 5 9 Spherical a-ray scintillation counter 1 X 1 0 ' 7 ~ 0 . 05

a-ray scintillation counter with electrostatic 1 X 1 0 " 8 ~ 0 . 0 3 5 sep aration

a-ray scintillation counter and adsorption of 5 x 1 0 - 9 ■ ~ 0 .0 0 1 radon from 100 1 expired air •

HURSH, J.B. [33] 1 9 6 3 Adsorption of radon on activated charcoal " < 0 . 01 (continuously) a-ray scintillation counter DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 27

essential advantage over the emanation method in that it is possible to local­ ize the individual radium deposit's. ' Estimation of the amount of radium deposited in the body of a poisoned patient by the measurement of the emitted gamma-ray radiation was per­ formed in the first cases of radium poisoning in the United States. Such investigations were first performèd by SCHLUNDT et al. [35] who used big ionization chambers switched in series with sensitive electrometers. The lower limit of detection was in the magnitude of some microcuries total-body radium, due to the imperfect insulation of the measuring electrode of the ionization chamber and to environmental radiation. In 1931/33 SCHLUNDT et al. [36] and IVES et al. [37] took into account the geometry of the human body and its self-absorption. They inserted sealed radon ampoules in a phantom and in this way imitated the real distribution pattern of radium in the human body. . . . ■ Modern development in the determination of incorporated radionuclides was introduced by EVANS, 1937 [9] and RAJEWSKY and FRANKE, 1938 [7, 39] who fo r the firs t time used a G eiger-M illler-tube counting device for the measurement of radium deposits (Fig. 4). The sensitivity was sufficiently good to satisfy the principal problem of the diagnosis of radium deposits, that is to say, the measurement' of radium amounts sm aller than 1 jug. Almost at the same time RAJEWSKY and DREBLOW studied the prob­ lem of determining the local distribution of radium deposits in the human body [7, 38] . They used a so-called gamma-ray stethoscope which enabled the diagnosing physician to contact directly the individual parts of the patient's body and to localize any radium deposit in this way (see Fig. 5). The sensitivity of this device was sufficient to detect amounts of radium as low as 0 .1 ¡je from a distance of 2 0 cm during a measuring time of about 2 0 min. • л. Some time before RAJEWSKY [7] and FRANKE [39] constructed an ionization chamber of 1 0 - 1 capacity which was able to detect radium deposits of 0.1 /лс (Fig. 6 ). An additional compensation apparatus rendered it possible to suppress the background due to environmental radiation still further. The apparatus was constructed in such a way that the chamber could be moved along three co-ordinates at right angles to each other and made it possible to localize the radium deposit in the patient's body. This was the first human counter installation. In 1947 HESS and McNIFF [40] succeeded in lowering the detection limit by a factor of 10 with a similar device. In this way radium deposits as low as 0.03 цс were detected in a period of 1 h; The first apparatus which rendered it possible to measure the gamma radiation of nuclides in normal concentration in the human body was built by SIEVERT about 1951 [41] (Fig. 7). SIEVERT and later BURCH and SPIERS [42] as well as RUNDO [43] were able to improve the sensitivity essentially. They measured the ionization current of several high-pressure ionization chambers arranged parallel to the prone patient. Another in­ crease in sensitivity was attained by reducing environmental radiation. Therefore SIEVERT et al. [44] built his apparatus in a mine. Other authors used water-filled shielding tanks. The arrangements described above' are subject to different disadvantages due to the employment of big high-pressure ionization chambers and water 28 В. RAJEWSKY et al.

Fig. 4

The gamma-ray quantum counter oí EVANS in 1 m arc position [9]

"Gammastrahlen Stethoscope" (7.38) tanks. This means bad economy of space and an endangering of patients and operators due to the employed pressures of 20-40 atm. Favoured by the development of scintillation substances and photomulti­ plier tubes ANDERSON changed from measurement of the ionization current DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 29

Localisation o f incorporated Ra22* by means o f a mobile 10-1 ionization chamber ( 7 . 39)

to the counting of gam m a-ray quanta, and developed a liquid scintillation counter [45] (Fig. 8). It consists of a cylindrical tank with a sm aller thin wall tube running along the axis. The annular space between the tube and tank wall is filled with the scintillation so ution. Scintillations are detected by 108 photomultiplier tubes inserted along the outer wall of the tank. The central cavity accommodates the subject or the sample within the sensitive shell so that the effective counting angle is nearly 4тг and the counting ef­ ficiency can be made nearly independent of source position. In the place of liquid scintillators BIRD and BURCH [46] employed plas­ tic scintillators for the first time and were able to remedy some technical difficulties on account of a good geometry of course (Fig. 9). Liquid or plastic scintillators as substances of low atomic number are subject to an essential disadvantage if it is desired to detect the presence of more than one radionuclide in the body. That is why MARINELLI [47] changed to an inorganic scintillator of high atomic number, thallium-activated sodium iodide (see Fig. 10). With reference to the three types of whole-body counters (Los Alamos Liquid Scintillator, Argonne Crystal Scintillator and Leeds Plastic Scin­ tillator), up to May, 1962 there were 86 counters developed and operated [48] ; 80% of them were crystal detectors. The possibility of the separation of gamma energies and the simultaneous determination of any deposits in the human body were the reasons why the Max Planck-Institute for Biophysics at Frankfurt/Main established a whole- body gamma-ray spectrometer "HUCO" similar to the counter used by MARINELLI and co-workers (Fig. 11) [47]. 30 В. RAJEWSKY et al.

Fig. 7

High-pressure ionization chamber-whole-body counter [41.44]

Table III (similar to Table II) is a historical review on the statements covering the lower limit of detection of different measuring arrangements which have been used in order to determine the total-body content of radium- 226 by its gamma radiation. Radium-226 is assumed to be in equilibrium with its daughter products.

2. LARGE SCINTILLATION DETECTOR REQUIREMENTS FOR WHOLE-BODY COUNTING

Estimation of the concentration of radionuclides in the human body re­ quires the measurement of very low activities. Requirements of systems able to detect such small activities are the following: the measuring period to be short; the measurement to be inde­ pendent of size and weight of the person to be tested; the apparatus to have an optimal resolving power and response; it should be possible to calibrate DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 31

Fig. 8

Los Alamos 4ir-liquid scintillation counter "HUMCO" [45]

Three plastic unit scintillation counter [46] the system by means of a phantom; no electronic drift over a long period; and the results to be reproducible. In the following sections some of these parameters will be briefly discussed. 32 В. RAJEWSKY et al.

F ig . 10

Argonne N al(Tl) whole-body scintillation counter in standard chair geometry [47]

Frankfurt Nal (TI) whole-body scintillation counter "HUCO” [49]

In general, it might be said, for any measurements of radioactivity, that the product of the total m easuring period times the square e rro r is only a minimum if "the resolution index" and the sensitivity are a maximum [52]. Sensitivity is defined here as the ratio of the measuring value to the activity of the sample, and the "resolution index" as the ratio of the square TABLE III EEMNTO O BD RDOCIIY HSOIA REVIEW HISTORICAL RADIOACTIVITY: BODY OF DETERMINATION DEVELOPMENT OF DEVICES FOR THE MEASUREMENT OF Ra™ TOTAL-BOD Y CONTENT BY WHOLE-BODY COUNTING SINCE 1929 AND ESTIMATION OF THE DETECTION LIMIT (ACCORDING TO [51])

• Approximate limit Duration of of detection A uthor (yr) D e v ice examination (or standard error) (¡ig Ra) (m in )

SCHLUNDT, H. et al. [35] 192!) ' Ionization chamber 5 ’ _ '

SCHLUNDT, H. et al. [21] 19 3 3 Ionization chamber 0 . 2 -

EV A N S. R. D . , [ 9 ] 19 3 7 G M -C o u n te r 0 . 1 - ■

RAJEWSKY. B. [7], FRANKE, I. [39] 19 3 7 10-1 ionization chamber in 0 . 1 10 compensation

RAJEWSKY, B. [7 ], DREBLOW, W. [38] 19 4 1 "Gammastrahlen-Stethoskop" 0 . 1 10

HESS, V. F. et al. [40] 19 4 7 Ionization chamber (1 atm .) 0 . 0 3 -

SIEVERT, R. M .- [41] 19 5 1 High pressure ionization chamber 0 . 0 0 5 SE 1 2 0

BURCH, P.R.J. et al. [42] 1 9 5 3 Differential high pressure ionization 0 . 0 0 3 SE 1 2 0 c h a m b e r

SIEVERT, R.M . [44] 1 9 5 6 High pressure ionization chamber 0 . 0 0 1 SE 1 8 0 - 2 4 0 (underground laboratory)

Argonne National Laboratory [47] 1 9 5 7 - Nal (Tl)-scintillation counter 0. 00035 SF 15 1 9 6 0

Los Alamos [45] 47Г liquid scintillation counter 0.0001 SE . 15

Leeds [50] • Plastic scintillation counter 0 . 0 0 0 1 5 SE 15

Frankfurt/Main [49] 1 9 5 9 / 6 0 N al(Tl)-scintillation counter 0 . 001 6 0 . 34 В. RAJEWSKY et al.

of the sensitivity to the background of the arrangement. Comparison of arrangements as to the values just mentioned is based on the presumption that there is an international agreement as to which radionuclides, which geometry and which carrier substances are to be used. In order to compare large scintillation detectors, the nuclides potassium-40 and caesium-137 are usually preferred. PFAU [53] discussed the total measuring period, the measuring effectiveness, the background and the maximum volume to be employed, and postulated the following result: If low activities are to be measured, the background should be reduced in order to increase the quality of the arrangement. The latter value means the reduction of the product of the total measuring period times the square of the mean square error of the measurement and a simultaneous increase of the ratio of the measured activity to the background. The background is due to the cosmic ray and its secondary radiation, as well as to the environmental radiation originating from internal and ex­ ternal sources (air and shielding material, etc. ). The background is re­ duced by mechanical shielding, air filtering and electrostatic filters and electronic devices such as energy discriminators and anticoincidence arrangements [54] . According to MEHL and RUNDO [48] 72% of all oper­ ating whole-body counters are iron- or steel-shielded. The most frequently used thickness of the shielding walls is in the range of 15-20 cm. Approxi­ mately 15% are lead-shielded and the remainder are shielded by water or chalk. Besides the parameters just discussed, the sensitivity of a large scintillation counter is dependent on its geometry and on the morphological composition of the person to be tested. This dependence may be reduced to a minimum, if the detector is arranged in a rotation symmetry around the patient's axis. For some detectors this implies some difficulties. If only a small sodium iodide crystal detector is used, the solid angle should be large enough so that the influence of the length dimensions of the patient is kept low. If different incorporated gamma-ray-emitting nuclides are to be detected, energy resolution should be high.

3. DISCUSSION OF THE PRINCIPAL TYPES OF WHOLE-BODY COUNTERS AND COMPARISON WITHIN THE PHYSICAL'PROPERTIES OF SELECTED DEVICES '

In Table IV whole-body detectors which are at present in common use are summarized. It includes high-pressure ionization chambers and scin­ tillation counters with inorganic and organic scintillators which measure gamma-ray and high-energy beta-ray radiation. The liquid scintillation counter is the most appropriate as to rapidity, reproducibility, efficiency and- simplicity of measurement. High-pressure ionization chambers are not very appropriate on account of their long measuring period. Besides, they have no energy resolution. The employment of inorganic scintillators has the advantage of good energy resolution; the geometry, however, ge­ nerally cannot be reproduced exactly. Another disadvantage is the smâll solid angle (efficiency = 0.2 to 0.5%) on account of the small crystal size. TABLE IV REVIEW HISTORICAL RADIOACTIVITY: BODY OF DETERMINATION

COMPARISON OF THREE DIFFERENT TYPES OF DETECTORS OF WHOLE-BODY COUNTERS [55]

High pressure ionization chamber Anorganic scintillators Organic scintillators

High energy High energy y в - ' 7 в ’ \ i ' ..- , i 1 P h o to - , C o m p to n Principle of Bremsstrahlung Photo-, Compton Bremsstrahlung effect measurement i e f f e c t - i ^ i Scintillation Ionization current j Photo-multiplier

Gas; N2 , СОг , Single crystal'. N al(Tl) p-terphenyl + POPOP in A ( 1 0 - 2 0 a tm . ) toluene or polyvinyltoluene Detector material PPO + 'POPOP in tri- ethylbenzene ,

y and high energy 8“ y (> 0. 01 MeV) y ( 0 . 1 - 3 . 0 M eV ) Detectable radiation 0(> 1.5 MeV) by its ' .6" (> 1. 5 MeV) by its • bremsstrahlung bremsstrahlung .

6 -1 0 % 18 - 30% Energy resolution - (K 40; 1 . 4 6 M eV ) (K40, 1.46 MeV)

Solid angle « 1Г 2 - 4тг

’ 10 - 22% (К 40) e.g . 0. 2% (K40) 12 - 25% (Со") E ffic ie n cy - 0.4% (Cs1*1) 6 - 8% (C s 137 ) ~ 0.3% (Sr90, bremsstrahlung)

CO СЛ со о>

TABLE IV (cont. )

High pressure ionization chamber Anorganic scintillators Organic scintillators AESY t al. et RAJEWSKY Approximate limit . 1 - 5 nc 0. 5 - 1 nc of detection % of MPB ' ~ i o ' 2 MPB (C s IS7, K40) (C s 137, K4 0 ) MPB (C o 60) = 10 (JC r. 226 Ra 1 - 5 x 10"4 0 .5 - lx 10'4 MPB Co60 M PB (S e 90) - 2 (jc MPB C o 60 ~ 5 x 10‘ 2 MPB Sr90 M PB (R aZZ6 ) - 0 , 1 fJc

Subject observations tim e* 3 - 4 h 2 0 1 0 0 m in 100 - 1000 s Standard error ± 50°/' ' < ± 10% ü ± 10% DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 37

Besides this, the measuring period is long compared with liquid scintillation counters. The employment of liquid scintillation counters is based on the utilization of the multiple Compton effect. Their energy resolution, there­ fore, is not as good as that of inorganic scintillation counters. Their measuring period, however, is reduced compared with inorganic scintillators due to th eir high e ffic ie n c y (up to about 7 0% if a 4jr geometry is used). Iñ Fig. 12 the counters most commonly used are shown schematically. On the left side of the figu re the c ry s ta l counters a re put together and on the right side the liquid or plastic scintillation counters are put together. According to MEHL and RUNDO [48] the standard chair geometry (Dl) is used with 55% of all whole-body counters. Each of the following: 1-2 m arc geometry (D2), stretcher geometry (D3) and geometry which can localize the incorporation simultaneously, are used at 20- 25%. ■ At'about 9% of the measurements 2v (D5) or 47r (D7,D8) geom etry is used, and at 2% the moder­ ated chair geom etry is used.

” 8 D-2 » IjVjl

D-3 ÈÙ

0 1 ■ д а д а

D-8

F i g . 12

Different detector devices with N ál(Tl) crystals (D1-D4) and organic plastic (D6) or liquid scintillators (DS, D7.and D8) [53] -

Whereas in the preceding sections the geom etric properties of the stan­ dard types of large scintillation counters were compared, in Table V this is done as to special types of whole-body counters. The background, sensi­ 38 В. RAJEWSKY et al.

tivity and the "resolution index" - as defined above - were normalized at a channel width of 80 keV in the energy interval of the photo-peak of potassium-40. If the normalized values are compared, it becomes evident that the Los Alam os type (a) and Genco type (b) have the highest, background (see line 9 in Table V). These arrangements, however, have the greatest sensitivity and "resolution index" (see lines 10 and 11,- Table V.) for the measurement of potassium-40. But if a gamma-spectrometric analysis of more than one nuclide is concerned, inorganic scintillation counters (such as the Argonne type (c) or the Harwell type (d)) are to be preferred to all the other arrangements on account of their much better energy resolution (see line 8 in Table V). Besides this, in Table V the physical properties of two inorganic scintillation counters of sim ilar type (as they are used, in Argonne and Frankfurt/Main) are compared. It is evident that both arrange­ ments are identical as to the data discussed.

' 4. CALIBRATION OF WHOLE-BODY COUNTERS

As to the in vivo measurement of incorporated radionuclides, it is ne­ cessary to distinguish between two different distribution patterns. Whereas potassium or caesium are practically homogeneously distributed within the human body, iodine is deposited in the thyroid gland and strontium or radium in the skeleton. Aerosol particles which are not soluble are deposited pre­ feren tially in the respiratory tract, and in the lung after inhalation. There­ fore it is necessary to imitate the kind of distribution pattern when calibrat­ ing the arrangement. As to cavity counters, such as the Los Alamos type or the Genco type with a 27Г or 4w geometry, the distribution of the incorporated radionuclide does not influence the sensitivity as much as it does with counters equipped with one or more single crystals. Consequently, the calibration of such counters is much easier. In principle, the best calibration method of whole-body counters is that of incorporating in human beings known activities of the nuclides to be tested. However, the incorporated nuclide is presumed to be equilibrated, that means to have a distribution similar to that in the steady state. During the period necessary for the equilibration process a portion of the radionuclide to be measured is excreted, and an activity estimate will be too low, especi­ ally for elements with a high equilibration period (e.g¿ t^e ^20 h). There­ fore, if the sensitivity of the counter is to be determined, the excreted portion should be known. Consequently, this calibration process is lim ited to elements with a long equilibration period. If, on the other hand, the phy­ sical half-life is small relative to the equilibration period, it is possible that after the physiological steady state is attained the activity would be below thé detection limit if the original incorporated activity was not in the order of magnitude of the maximum permissible amount. In cases whera calibration by incorporation is not possible, the calibration should be perform ed by means of a phantom. If a quasi-homogeneous distri­ bution pattern of the incorporated radionuclide is to be assumed, the radio­ nuclide concerned is dissolved in water - presenting a material similar to tissue - and sealed in cylindrical plastic containers. This method will TABLE V REVIEW HISTORICAL RADIOACTIVITY: BODY OF DETERMINATION

COMPARISON OF SIX DIFFERENT WHOLE-BODY COUNTERS (ACCORDING TO [53])

Los Alamos (a) G en co (b) Argonne (c) Harwell (d) L e e d s (e ) Frankfurt/Main (f)

Solid angle . ~ 4 tt — 2tt < 7Г < 2тг < 2тг < ÏÏ

Position of the patient Transport sleigh Patient standing Standard chair S tre tc h e r M od. ch a ir 1 m a rc stre tch e r

Detector material T e rp h e n y l- T erp h en y l- N a l(T l) N al (T l) O rg an ic N a l(T l) POPOP-PPO POPOP scintillators (Shell oil)

Detector dimensions 1 cylindrical 1 half-cylindrical 1 cylindrical 4 cylindrical 3 plastic ' 1 c y lin d ric a l tank, 1600 1, tank, 280 1, c ry sta l cry sta ls scintillators cry s ta l 30 cm thickness 15 cm thickness 20 cm x 10 cm 0 11 cm x 5 cm 0 50 x 25 x 17 cm 20 cm x 10 cm 0

Background in K40 3 5 0 130 3 . 5 0 . 9 17 1 . 6 énergy interval (Rn) [c p s ]

Efficiency (E) for K40 . 68 12 0 . 7 0 . 2 3 . 9 0 . 4 • phantom measurement (5 0 kg)

x 100% 1 [y P s J

Ez/Rn in K40 energy 1 3 . 4 1 .1 0 . 14 0 . 04 0 . 8 5 0 . 1 in te rv a l

Energy interval for K40 1.15 - 1.55 0.94 - 1.74 1. 26 - 1.66 1 . 4 2 - 1 . 5 0 1.10 - 1. 58 1.29 - 1.63 [M e V ]

GO CD о

TABLE V (cont.)

Los Alamos (a) G en co (b) Argonne (c) Harwell (d) . L e e d s (e ) Frankfurt/Main (f)

Energy resolution °}o 3 4 5 0 6 6 27 8 . 7 for К 40

Normalized background 70 13 0 . 7 0 . 9 2 . 8 0 . 4 al. et RAJEWSKY. in K;40 e n erg y in te rv a l [cps/80 keV]

Normalized counting • 1 3 . 6 1 . 2 0 . 14 0 . 2 0 . 6 5 0 . 1 efficiency for K40 p er 8 0 keV Ш

E2/Rn for normalized ' 2 . 7 0. 11 0 . 0 3 0 . 04 0 . 1 4 0. 0 2 5 counting efficiency (K40)

Estimated total price 240 000 DM - 130 000 DM - 220 000 DM - 190 000 DM - 180 000 DM - - 250 000 DM . for installation 4 0 0 0 0 0 DM 1 6 0 0 0 0 DM 2 8 0 0 0 0 DM 4 0 0 0 0 0 DM 2 2 0 0 0 0 DM DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 41

F i g .13

Comparison of relative counting rates for a phantom filled with Cs137 in solution ' and for two persons contaminated with Cs131 [56] о Man, 60 kg, 165 cm; x Man, 81 kg, 187 cm; - Phantom

— + MEASURED DISTRIBUTION (CASE FE )

— о CALCULATED DISTRIBUTION (TOTAL) (THOROTRAST WITHIN LIVER/SPLEEN ~ 1 2 0 0 ANO NECK) LU — • CALCULATED DISTRIBUTION (THOMAS U) 5 1000 (THOROTRAST WITHIN THE NECK) E — X CALCULATED DISTRIBUTION (THOMAS I) £ 800- ( THOROTRAST WITHIN LIVER AND SPLEEN)

ÜJ 5 600- £E . ^ 4 0 0 ­ P PARAV. DEPOT (NECK) § 200 ■ 1 ^ О X------* 0 0 10 20 30 «0 60 60 70 B0 90 100 120 cm THE PATIENT (+ ) POSITION OF THE CRYSTAL ALONG THE AXIS OF THE PHANTOMS (•« )

F ig . 14

Comparison of counting rates for a phantom with liver, spleen and perivascular Th228 deposit • on neck and for a thorotrast patient [49] do in most cases. RUNDO [56] considered how far this mode of action is a true imitation of in vivo conditions. For a certain geometry he compared the normalized counting rates of the phantom arrangement and the equivalent measurement of a person who had incorporated a nuclide to be considered. Both counting rates were plotted against the relative position of the crystal and phantom, or person under test, respectively (see Fig. 13). If the distribution pattern is inhomogeneous, however, it is necessary to imitate within the phantom the distribution pattern observed in vivo where the position of the crystal along the axis of the patient was varied. The com­ parison of the results (Fig. 14) reveals a calibration of a single crystal scin­ tillation counter to be possible even if an inhomogeneous distribution pattern is involved [49] . Investigations were performed in order to decide whether or not the results obtained from potassium-40 phantom measurements are still valid rfb to

TA B LE VI

COMPARISON OF PHANTOM AND IN VIVO-CALIBRATION OF A WHOLE-BODY COUNTER FOR K 40 [61]

Energy interval Total efficiency Calibration Isoto p e Incorporated amount or activity Calibration factor • ДЕ (M eV ) (%)

CO RJWK e al. et RAJEWSKY. P h an to m K n at 1 . 2 9 10029.818 g KC1 1 cp m ■

■ - 1 . 6 3 - 5285.71 g Knat 1.2 g i 0.012 g Knat 0. 40 ± 0. 004

P h an tom К 42 ' 1 . 3 6 1 cp m = 1 2 . 0 ¿к: - 1 . 7 0 0. 52 ± 0. 005 ne K42 0.40 i 0. 004

In vivo К 42 1 - 3 6 1 cp m 72.2 - 118.6 jic (6 persons) - 1 . 7 0 - ’ 0.49 ± 0. 05 ne K42 ' 0. 38 ± 0. 04 DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 43

for in vivo measurements. The counting rate of the phantom equipped with plastic bottles containing an aquéous potassium-42, solution was compared with the counting rate obtained from a test person having incorporated potassium-42. The calibration for potassium-40 by means of potassium-42 is possible, as both isotopes emit a nearly identical gamma-ray energy at 1.46 and 1.53 M eV. G REEN and M cN E IL [57] found the deviation between the two calibration methods to be approximately 4%. Table VI reveals the calibration factors of phantom and in vivo calibration to be identical.

5. EMPLOYMENT OF WHOLE-BODY COUNTERS

If any radioactivity is to be estimated in the human body, the following questions are of interest: the nature of the nuclide, its localization, its stay­ ing period within the organism and the chemical process to which it is sub­ ject in the organism. . These questions may be answered by a large gamma-ray counter so that the scope of-its employment implies the following field of investigation; esti­ mation of the content of naturally-occurring radionuclides in the human or­ ganism, medical diagnosis, biochemistry, studies of metabolism and de­ termination of the content of artificial radionuclides caused by incorporation.

5. 1. Estimation of the normal content of radionuclides occurring in the human organism

■ The radionuclides occurring in the human organism include not only the naturally-occurring radioelements but also the radionuclides due to atomic weapon tests via the biocyclus. In Table VII the most important gamma-ray-emitting radionuclides occurring in the human organism, and their total-body activity are summarized. From these scheduled radio­ nuclides only the naturally-occurring potassium-40 and the artificial caesium- 137 are detectable by means of their gamma-ray spectrum. Investigations on the natural radioactivity of the liying human being were performed for the first time by SIEVERT [41, 44], BURCH and SPIERS [42] and later on by RUNDO [43], who determined the total-body content of po­ tassium by means of ionization chamber counters. The results obtained in this way are in good agreement with the content determined by a normal chemical analysis as demonstrated in Table VIII. ANDERSON [62] found for the first time by whole-body counting that the content of potassium-40 is a function of age in man. The results were confirmed later by ONSTEAD et al. [60]. . ' In 1956 MILLER and MARINELLI [63, 64] detected caesium-137 in the human organism for the first time. The content of caesium-137 is subject to temporal and local variations. This fact is demonstrated by the specific activities found by M ILLE R and M AR IN E LLI [65] who determined the content of caesium-137 in the period from April 1956 to July 1958 with eight test persons in relation to their total-body potassium content as follows:

April 1956 33 pc caesium-137/gK June 1957 36 pc caesium -137/gK 44 В. RAJEWSKY et al.

April 1958 47 pc caesium-137/gK June 1958 52 pc caesium-137/gK

The average values of the total population living in the United States were the following:

1956 41 pc caesium-137/gK 1957 44 pc caesium -137/gK 1958 54 pc caesium-137/gK

TABLE VII

AVERAGE NORMAL TOTAL-BODY CONTENT OF DIFFERENT RADIONUCLIDES

N uclide Average total-body content V

. (PC)

' Potassium-40 < 10 s

R ad iu m -226 4 6 *

L e a d -210 4 6 0 *

Thorium-228 1 8 * * ' -

Caesium-137 1 2 0 0 * * *

* see [58,59] * * 20% of Th228 activity assumed to be in soft tissue [58.59] ** * Average Cs137 content of a normal unexposed German person in ' 1959/60 [60]

ONSTEAD et al. [60] observed a reduction in the content of caesium-137 in the human organism beginning in the spring of I960. This runs parallel to a reduction of contamination of the biosphere due to the cessation of atomic weapon tests. The content of the other artificial radionuclides in the human organism is subject to sim ilar variations due to variations of the degree of contamination of the biosphere. The content is, however, often below the detection limit; that is why this subject is covered by a few publications only. . In 1959 NIELSON [67] found the content of zinc-65 to be between thé limits of 2 and 36 nc. The average of 172 measurements performed in 1959 at Hanford Laboratory [ 6 8 ] was 3.4 nc in the total body. RUNDO and NEWTON [69] detected zirconium/niobium-95 in 15 persons being measured. As these findings were suspected to be due to a retention of radioactive particles in the lung, the measuring arrangement was cali­ brated by means of a breast phantom. The following measurement of persons to be tested resulted in a specific activity as high as '0. 35 pc zirconium/niobium-95 per gram of lung tissue, being equivalent to an ab­ sorbed d ose-rate as high as, 0.107 /jrad/h. ' DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW

TABLE VIII

THE AVERAGE TOTAL-BODY POTASSIUM CONTENT BY . INDIRECT (CHEMICAL AND ISOTOPE DILUTION) AND DIRECT ANALYSIS (WHOLE-BODY COUNTING) . (ACCORDING TO [44, 61]) ,

A verage Specific A ge Author (yt) (N o .) sex body mass K -con ten t (yr) (kg) (g/k g )

SHOHL, A .T. + 1939 - ^ - - - 2 . 1 s

* HAWK, P.B. et al. * 1947 - - - - 3 . 5 0

CORSA, L. et al.++ 1950 30 m 2 1 -3 2 - 1.8i ± 0.19

AIKAWA, J. W. et a l.++ 1952 30 f ' - - 1. 2j ± 0 .1

EDELMANN, J. S. 1952 33 m 1 .8 o e t a l. ++ - - 14 f 1 . 6 0

BLAINEY, J.D . etal.++ 1954 17 m 1 . 7 , ± 0 .1 4

7 f 1.3, ± 0.19

BURCH, P.R.J. et al. • 195 4 10 m -

1 9 -2 0 - 2 . 1 s ± 0 . 1 3 f

11 m 2 0 -4 1 - 2 . 1 2 ± 0 .1

4 m 6 0 -7 9 . - 2. 15 ± 0 .2

RUNDO, J. et al. ++ 19 5 5 6 m 2 2 -3 1 76 2 . 15 ± 1 .2

4 f 1 7 -2 8 61 1 . 9 7 ± 1 .5

SIEVERT, R.M. et al. 1 9 5 6 /5 7 20 m 1 0 -1 3 40 1 . 9 i ± 0 .0 8

11 m 2 0 -2 9 70 1 . 9 6 ± O .lo

10 Ш 3 0 - 4 9 70 2 .0 , i 0 . 1 3

14 m 6 2 - 8 4 . 70 ; 1 . 5 5 ± 0 .0 8

17 f 1 0 -1 3 40 1 . 6 4 ± 0 . 1 ,

4 f 2 5 -2 8 70 ' 1 . 4 S ± 0 .0 5

5 f 3 6 -5 6 70 1.3, ± 0.09

18 f 66-86 70 1. 2 , = 0 .0 8

MARINELLI, L. D. 1956 12 m 2 2 - 3 4 - 1. 8 g ± 0 . 06

3 f 2 2 - 2 9 - 1 . 5 4 ± 0 .0 3 46 В. RAJEWSKY et a l.

TABLE VIII (cont. )

A verage Specific Age Author (yr) (N o .) sex body masb K -con ten t (yr) (kg) (g /k g )

ANDERSON, E.C . ' 1956 28 m - - 2 . 1 ,

16 f - - 1 . 6 0

OWEN, R .B . 1957 - - - - 2 . 5 * '

RUNDO, J. 1957 10 m - . 70 1 .7 ± 0 .0 9

McNEILL, K.G. et al. 1959 30 m ~ 2 0 79 2 . 1 2 ± 0. 2 3 .

ONSTEAD, C.O. 1960 - - 3 0 -4 0 - ' 2 .1 e t a l. - - . - ’ 1 .6 ■

BOULENGER, R.R. 1961 103 m .f - - 2 . 0 „ ± 0.2 „ e t a l.

REMENCHICK, A.P. 1961 28 m , f 2 8 - 7 3 83 1 . 8 , 1 0 . 3 7 e t a l.

DELWAIDE, P. A. 1961 " 10 m - - l . ? 6 i 0 . 6 , e t a l.

FORBES, G. В. et al. 1961 42 m 1 1 -4 4 22 1.36 - 2,26*

8 f 7 - 2 3 - 1 0 5 0.9 - 2.02*

KAUL, A. et al. 1 9 6 2 /6 3 148 m 1 5 -6 0 73 1 . 6 , 1 0 .2 s

37 f 2 0 -4 0 60 1 . 5 ! ± 0 . 2 3

1.9, i 0.18** i 9

* Not considered for the average * * Standard error + Chemical analysis ++ Exchangeable potassium DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 47

Findings on iodine-131 are published by RUNDO [70] and MUTH et al. [71]. • . ■ RUNDO [70]- and V A N D IL L A et a l. [72] reported on strontium -90 in­ corporation. They measured the bremsstrahlung of the betas being emitted.

5. 2. Nucleai—medical diagnostic and biochemical-medical research

HEINRICH and PFAU [55] reported on investigations covering intestinal resorption, biological half-life and the turnover rate of vitamin B12. Cobalt, the central atom of this compound, was marked by its cobalt-60 isotope counted by large liquid scintillation counters which render it possible to per­ form studies over a long period with tra cer amounts sm aller than the natural activity in the body. Thus the body burden is kept to a minimum. This ad­ vantage is also described by REIZENSTEIN et al. [73] . HEINRICH and PFAU [55] were successful in reducing the usual activity of vitamin B 12 m arked with cobalt-60 which is employed in the clinical radio-vitamin B12 diagnosis from 0 .2 or 0 .1 pc to 0 . 0 1 juc. - Investigations on the metabolism of potassium, using whole-body count­ ing and simultaneous flame-photometric determinations, opened a discussion as to how fa r a reduction of the total-body content of potassium in the healthy man is accompanied by a potassium reduction in the erythrocytes, the plas­ ma and the residual intracellular space [58, 74, 7 5] . The turnover of different radionuclides in the organism is published in the literature with diverse details. HINE et al. [76] reported on investi­ gations covering the metabolism of sodium after injection of a dose of sodium - 2 2 ranging between 1 0 and 2 0 /не. MALAMOS et al. [77] estimated the resorption of iron-59 in the gastro­ intestinal tract seven days after oral administration. The same authors re­ ported on the intestinal resorption of vitamin B 12 labelled with cobalt-60, the turnover rate of proteins being studied by means of intravenously-injected human serum albumin marked with iodine-131, and measurements of bromine-82. SARGENT [78] subjected patients to irradiation with 910 MeV alpha particles accelerated by means of a cyclotron. In the body, carbon-11 was formed by a (а, cm) process out of carbon-12. SARGENT was able to detect carbon-11 in the expired air and by whole-body counting. He also investi­ gated the oral resorption and long-period retention of iron-59 in the bodies of patients suffering, from a pathological iron metabolism. Besides this he studied the turnover rate of calcium-47 in the body of a patient suffering from acromegalia. The rate of turnover of albumin and globulin marked with iodine-131 in the body of patients suffering from neoplastic diseases was estimated by COHN et al. [79]. These authors also determined the resorption and turn­ over rate of vitamin B12 marked with cobalt-58, which was applied topatients suffering from anaemia or poor intestinal resorption. They also studied different parameters influencing the metabolism in bones under pathological conditions. To-date strontium-89 or calcium-47 was used as a tracer. Simi­ lar investigations covering tumours in bones and métastasés in the- skeleton were described by CEDERQUIST and LIDÉN [80]. These authors reported also on the retention of iodine-131 under normal and pathological conditions. 48 В. RAJEWSKY et al.

The metabolism of thorium-232 and its daughters was investigated by­ several teams. They examined patients in whom thorium dioxide was injected intravenously for diagnostic reasons as a contrast medium for X-rays. The physiological steady state of thorium and its daughters was estimated by simultaneous whole-body measurements and determination of excreted thorium and decay products [49, 84, 85, 86, 87, 88].

5. 3. The employment of whole-body counters in the case of radiation accidents '

In 1958 a radiation accident occurred at Oak Ridge in the United States and eight persons were exposed to neutron and gamma radiation. In vivo measurements were performed with a whole-body scintillation counter equipped with a thallium-activated sodium iodide crystal ip order to estimate the neutron dose absorbed by these persons. The estimation was performed by measuring the sodium-24 activity in the energy range of 2.7 5 MeV which was formed by a neutron activation process out of the stable sodium in the body [89, 90]. . MARINELLI [91] reported on a radiation accident caused by incorpo­ ration of americium-241 which could be detected in the body of the patient by gamma-ray.spectrometry. In another case the author detected tellurium- 129 which was incorporated by inhaling particles of an aerosol carrying tellurium-129. MacDONALD [92] detected the incorporation of antimony-124 as well as magnesium-54 in men operating reactors. The latter isotope was formed by neutron activation and was inhaled with dust.

6 . DISCUSSION OF THE LOWER LIMIT OF DETECTION OF RADIONUCLIDES BEING MEASURED IN VIVO BY WHOLE-BODY , COUNTERS IN REGARD TO THE MAXIMUM PERMISSIBLE BODY . BURDEN . ■

In Table IX the detectable amount of some of the radionuclides in re­ lation to the maximum permissible amount is scheduled [5, 48] . As already mentioned above only a few of the radionuclides naturally present in the human organism have a concentration above the detectable level. So it be­ came necessary to lower the limit of detection by new measuring methods. The sensitivity attained up to now is not sufficient to determine, e.g. the normal total-body radium-226 content in man. Besides this, the detection of some radionuclides emitting low-energy beta-radiation is still difficult. This is true for the isotopes strontium-90 and plutonium-239 which are very important from the radiobiological point of view. On the other hand, many radionuclides can be estimated by the employ­ ment of low-level whole-body scintillation counters with a detection limit of the order of or far below the maximum permissible body burden. In ad­ dition, it became possible to perform clinical tracer tests employing acti­ vities far below the maximum permissible level, so that the patient's body burden could be reduced. DETERMINATION OF BODY RADIOACTIVITY: HISTORICAL REVIEW 49

TABLE IX

DETECTION LIMITS OF WHOLE-BODY COUNTERS IN FRACTIONS ■ OF THE MAXIMUM PERMISSIBLE BODY BURDEN [48]

1 0 's - 10 " 4 (М РВ) 10"* - 1<Г3 (МРВ) 1 0 '3 - 10" (MPB) • 1 0 '2 - 10 (MPB) 10'1 - 10° (MPB)

N uclide МРВ (р с ) N u clide MPB (fic) N u clide MPB (p c ) N u clide MPB (itc ) N u clide MPB (p c )

Be7 600 . N a2’ 10 M n56 10 Sr/Y9° 20 Pb210 4 / p l8 20 N a24 7 N i65 10 I13! (D ) 20 Ra223 (D ) 0.07

Sc47 20 К 42 10 Z n ram 30 H o166 30 A c 228 (D ) 0.09

C r51 800 С а47 (D ) 10 Nb97 20 Pb212 (D ) 0.2 T h 227 (D ) 0.1

M n54 40 Sc46 20 Ru/Rh106 10 Ra224 (D ) 0.07 Np237 (D ) 0.5

С о я 200 Sc48 9 T e 129m 20 Ra226 (D ) 0.2 C m 243 (D) 0 .3 . T e 129 jl2 6 .C o 58 30 V48 10 90 Ra228 (D ) 0. 09 C m 245 0.4

jia i Z n65 60 M n52 9 50 Th22" (D ) 0.09

jl32 A s 74 40 Fe59 20 10 U 2SS 0 .4

jl33 Sr85 60 С о60 10 20 A m 241 0.3

Z r95 (D ) 20 C u64 80 Ba140 (D ) 9 A m 243 (D ) 0 .4

N b95 40 G a72 10 C e 144 (D ) 20 ' C f249 0.3

Ru103 50 A s76 20 Pr142 20-

Inll3m 70 A s77 80 N d147 50

Sb125 60 Br82 10 T m 180 60

Cs134 20 Rb86 30 T a 182 20

Cs136 30 Zr97 (D ) 9 Re188 20

A u 196 40 Rh105 100 B i2(ri/z(r7mpb 20

A giio/nom A u 198 30 10

T I202 50 Sb122 20

Sb124 10

РЬШ 90 T e 132 (D ) 10

Cs137 30

Ра233 60 La140 10

C e 141 90

Eu154 20

G d159 50

T b 160 20

Er1T1 30

Y b 175 100

Re186 50 ■* Ir192 20

Hg2ie 80

Np239 70

Assumptions (H.G. MEHL, J. RUNDO £48]): ' . '

(a) A. Single crystal, 20 cm diam. by 10 cm thick; (b) Subject activity and counter background are measured for 20 min each; (c ) Only one contaminating nuclide of known identity is present in addition to 140 g К and 10 nc Cs137; (d) The detection lim it is three times the statistical standard error on the net counting rate in the photopeak region; (e) Nuclides which have gamma-ray-emitting daughters have the indication (D) 5P В. RAJEWSKY et al.

REFERENCES

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[47] MARINELLI, L. D ., Brit. J. Radiol., Suppl. 7 (1957) 38. ' [48] MEHL, H.G. and RUNDO, i., Health Physics 9 (1963) 607. ■ . • [49] KAUL, A. and RAJEWSKY, В ., -Proc. 2nd Symp. on Radioactivity in Man, Chicago/Ill. (1962), in press. [50] BURCH, P .R .J., HUGHES, D. , IINUMA, T.A. , OVERTON, T.R. and APPLEBY, D. B. , in Whole-body counting (IAEA-Symp. , Vienna, 13-16 June 1961) 59. . [51] SPIERS, F. W. , "Whole-body counting: and introductory review", in: Whole-body counting, Proc. Symp., Vienna (June 1961), IAEA, Vienna (1962) 3. . [52] LOEVINGER, R. and BERMAN, М ., Nucleonics 6 (1951) 26. [53] PFAU, A. , Atomwirtschaft 8 (1963) 433. [54] MILLER, C.E. , MAY, H.A. and MARINELLI, L.D ., in: Health.Phys. in Nucl. Install. (OEEC, ENEA, . DANISH A EC Eds) RiseS Symp. (1959) 119. ' [55] HEINRICH, H.C. and PFAU, A ., Atomkemenergie 6 (1961) 463. [56] RUNDO, J. , "Some calibration problems of whole-body gamma-ray spectrometers”, in: Whole-body counting, Proc. Symp. , Vienna (June 1961) IAEA, Vienna (1962) 121. [57] GREEN, R.M. and NcNEILL, K .G ., Austr. J. Appl. Sci. 13 (1962) 66. [58] UNITED NATIONS Scientific Committee on the Effects of Atomic Radiation Report, Suppl. 16 (A 15216) United Nations, New York (1962). [59] STAHLHOFEN, W ., Thesis, Frankfurt/Main, in preparation. [60] ONSTEAD, C .O ., OBERHAUSEN, E. and KEARY, F. V ., Atompraxis 6 (1960) 337. [61] KAUL, A., SCHOEPPE, W ., KOCH, K.M. and H1ERHOLZER, K ., Biophysik (1964), in press. [62] ANDERSON, E.C. and LANGHAM, W .H ., Science 130 (1959) 713. [63] MILLER, C.E. and MARINELLI, L .D ., Science 124 (1956) 122. [64] MILLER, C.E. and MARINELLI, L.D. , ANL-5518 (1956) 52. [65] MILLER, C.E. and MARINELLI, L. D., ANL-5919 (1958) 74. [66] LANGHAM, W.H. and ANDERSON, E .C ., Health Physics 2 (1959) 30. [67] NIELSON, J.M . , Science 129 (1959) 94, [68] ROESCH, W .C ., McCALL, R.C. and PALMER, H .E., HW - 67045 (1960) 33. [69] RUNDO, J. and NEWTON; D., Nature 195 (1962) 851. . [70] RUNDO, J., in: Proc. 2nd UN Int. Conf. PUAE 23 (1958) 101. [71] MUTH, H. and OBERHAUSEN, E. , Sttahlentherapie 120 (1963) 550. [72] VAN DILLA, M. A. and ANDERSON, E.C. , United States Armed Forces M .J. 11 (1960) 526. [73] REIZENSTEIN. P.G. , CRONKITE, E.P. and COHN, S.H. , Blood 18 (1961) 95. [74] SCHOEPPE, W ., KAUL, A ., KOCH, K.M. and HIERHOLZER, K ., Klin. Wochenschr. 41 (1963) 463. . [75] SCHOEPPE, W ., KOCH, K .M ., KAUL, A. and HIERHOLZER, K ., Proc. Int. Congr. Endocrinol. Nucl, M ed., Freiburg, (1963), in press. [76] HIÑE, G. J. , JAGGER, P. J. and BURROWS, B. A ., "Use of a clinical body counter for long-term ex­ changeable sodium studies", in: Whole-body counting, Proc. Symp., Vienna (June 1961) IAEA, Vienna (1962) 413. . [77] MALAMOS, B ., BELCHER, E. H. , BINOPOULOS, D. and CONSTANTINIDES, C. , "The use of whole-body

counting techniques in clinical studies with FeM, Na22, Br82, Coss-labelled Vitamin B12 and H3i-labelled albumin", in: Whole-body counting, Proc. Symp., Vienna (June 1961) IAEA, Vienna (1962) 43&. [78] SARGENT, T.W . , "Metabolic studies with Fe59, Ca4^ and Си in various diseases", in: Whole-body counting, Proc. Symp. , Vienna (June 1961), IAEA, Vienna (1962) 447. [79] COHN, S.H ., LIPPINCOTT, S.W ., CRONKITE, E.P. and REIZENSTEIN, P.G. , "Application of whole- body gamma spectrometry to clinical tracer studies", in: Whole-body counting, Proc. Symp., Vienna (June 1961), IAEA, Vienna (1962) 469. [80] CERDERQUIST, E. S. and LIDÉN, K .V .H ., "Clinical use of a Nal(Tl) body counter”, in: Whole-body counting, Proc. Symp., Vienna (June 1961), IAEA, Vienna (1962) 487. [81] RUNDO, J ., Brit. J. Radiol. 28 (1955) 615. [82] RUNDO, J. and FABER, M. , Recueil de Travaux Chimiques des Pays-Bas T 74 (1955) 416. [83] HURSH, J.B ., STEADMAN, L .T ., LOONEY, W.B. and COLODZIN; M ., Acta Radiol. 46 (1956) 481. [84] RUNDO, J ., Acta Radiol. 47 (1957) 65. [85] MILLER, C .E ., ANL-5829 (1957) 144. [86] RUNDO, J., Thesis, London (1958). [87] MUTH, H. and OBERHAUSEN, E ., "Measurement of radium and thorium burdens", in: Whole-body counting, Proc. Symp., Vienna (June 1961), IAEA, Vienna (1962) 267. 52 B. RAJEWSKY et al.

[ 8 8 j STAHLHOFEN, W. and KAUL, A ., "Measurement of Ra226, Th23Z and daughter products in biological samples", in: Proc. Symp. on Radiol. Health and Safety in Mining and Milling of Nuclear Materials, II_ (Aug. 1963), IAEA, Vienna (1964)475. [89] COFIELD, .R.E. , Y-1283. [90] Report Y-1234. [91] MARINELLI, L .D ., Minerva Nucleare 1 (1957) 201, [92] MacDONALD, N .S ., "Recent uses of a total-body counter facility in metabolic research and clinical diagnosis with radionuclides", in: Whole-body counting, Proc. Symp., Vienna (June 1961), IAEA, Vienna (1962) 501. IN VIVO COUNTING (Sessions 2 and 3)

A TWO-CRYSTAL SCANNING-BED COUNTER FOR ACCURATE DETERMINATION OF WHOLE-BODY ACTIVITY

■ Y. NAVERSTEN RADIATION PHYSICS DEPARTMENT, UNIVERSITY OF LUND, SWEDEN

Abstract — Résumé — Аннотация — Resumen

A TWO-CRYSTAL SCANNING-BED COUNTER FOR ACCURATE DETERMINATION OF WHOLE-BODY ACTIVITY. Whole-body counters often utilize a detector in a chair arrangement. Such a geometry has also

been used at our laboratory for potassium and Cs*37 measurements. However, for several elements or compounds redistributions within the body make this geometry not very well suited for whole-body activity determinations. A geometry with two scanning detectors, one above and one below a stretcher, has been developed for the whole-body counters at Lund. The efficiency of this geometry with 5in x 4in Nal(Tl) crystals has proved to be very insensitive to the distribution pattern of the activity within the body. This paper gives iso-response curves for point sources in the 144/45-cm two-crystal scanning bed geo­ metry, when no absorption losses are considered. The effect of absorption in water is given for a simple case. The effect of redistributions of psi, Cs*3* and Ca*7 is studied by means of an Alderson phantom counted in both the 42-cm chair geometry and the 144/45-cm scanning bed geometry. An example from the clinical work with Sr85 on patients is also given. Experience of the scatining-bed geometry shows that the whole-body activity may be assessed to within 15% accuracy, regardless of the distribution in the body. Improved scanning geometries with the patient lying on an arc, or geometries with two crystals moving in arc orbits are suggested.

UN ANTHROPOGAMMAMÈTRE DE PRÉCISION A DEUX CRISTAUX. Dans beaucoup d’anthropogamma- mètres, le sujet est examiné en position assise. Cette géométrie a été également choisie au laboratoire de

l’auteur pour les mesures de K et de 137 Cs. Toutefois, pour certains éléments ou composés, des redistributions

à l’intérieur de 1‘organisme rendent cette géométrie assez mal adaptée aux dosages de l'activité du corps humain. Pour les anthropogammamètres, on a mis au point à Lund une géométrie comportant deux détecteurs, l'un au-dessus, l’autre au-dessous d'un brancard. On a constaté que l'efficacité de cette géométrie, avec des cristaux de Nal(Tl) de 12,5 xlO cm , n'était guère influencée par le schéma de répartition de l’activité dans l'organisme. . L'auteur donne des courbe iso-réponses pour des sources ponctuelles dans la géométrie de la gamma- graphie en position couchée avec deux cristaux (144/45 cm ), en supposant qu'il n’y a aucune perte par ab­ sorption. Il indique l'effet de l’absorption dans l'eau pour un cas simple. Il étudie l’effet des redistributions

d e l3LI, L32Cs et Ca au moyen d*un fantôme d* Alderson, qui est compté aussi bien dans la géométrie en position assise (42 cm) que dans celle de la gammagraphie en position couchée (144/45 cm ). Il donne égale­

ment un exemple emprunté à des travaux cliniques sur des malades à l'aide de 8 5 Sr. L’expérience acquise dans la géométrie de la gammagraphie en position couchée montre que, quelle que soit sa distribution dans l’organisme, l'activité du corps peut être déterminée à 15% près. L'auteur propose des géométries améliorées, -dans lesquelles le sujet est couché en arc, ou dans lesquelles deux cristaux se déplacent sur des orbites en arc de cercle. '

СЧЕТНОЕ СКЕННИРУЮЩЕЕ УСТРОЙСТВО С ДВУМЯ КРИСТАЛЛАМИ ДЛЯ ТОЧНОГО ОПРЕДЕЛЕНИЯ СОДЕРЖАНИЯ ИЗОТОПОВ В ОРГАНИЗМЕ. В счетчиках для измерения радиоактивности всего организма детекторы часто располагаются в геометрии кресла. По­ добная геометрия использовалась и в данной лаборатории для измерений содержания калия и C s137. Однако при перераспределении в организме некоторых элементов или соединений эта геометрия не очень удобна для измерения активности всего организма. Для счетчика Лунде была разработана геометрия с двумя скеннирующими детекторами, расположенными над и под носилками. Эффективность этой геометрии с кристаллами 12,5 X 10 см NaJ(Tl) оказалась нечувствительной к виду распределения активности в организме. В настоящем докладе даются кривые изо-чувствительности для точечных источников с двумя кристаллами в скеннирующей геометрии койки 144/45 см, когда не наблюдается потерь

55 56 Y . NAVERSTEN

поглощения. Эффект поглощения в воде дается для простого случая. Эффект перераспре­ деления иодя-131, цезия-132 и кальция-47 изучается посредством фантома Альдерсона, высчи­ танного как с помощью геометрии42-см кресла, так и скеннируюшей геометрии койки 142/45 см. Приводится также пример из клинической работы с пациекгами с использованием стронция-85. Опыт скеннирующей геометрии койки показывает, что активность всего организма можно определить с точностью до 15% независимо от распределения в организме. Предлагаются улучшенные геометрии скеннирования на пациентах, лежащих на дуге, или геометрии с двумя кристаллами, перемещающимися по дуге.

ANTROPOGAMMÁMETRO DE PRECISIÓN CON CONTADOR DE DOS CRISTALES Y CAMILLA DE EX­ PLORACIÓN. Los antropogammámetros suelen comprender un detector y una silla en la que se. instala el sujeto. Esta geometría ha sido utilizada en el laboratorio del autor para efectuar determinaciones de potasio y de 137 Cs. Sin embargo, cuando se trata de estudiar la redistribución de varios elementos o compuestos en el organismo, dicha geometría no es la que más se presta para determinar la actividad del cuerpo entero. Por tal motivo, se ha ideado para los antropogammámetros del laboratorio de Lund otra geometría, con dos detectores de exploración, situados uno encima y otro debajo de una cam illa. Se ha comprobado que el rendimiento de esta geometría, con cristales de Nal(Tl) de 12,5 x 10 cm no depende del esquema de dis­ tribución de la actividad en el organismo. La memoria da las curvas de isorrespuesta a fuentes puntiformes en una geometría de cam illa y contador de exploración de dos cristales de 144/45 cm , sin tener en cuenta las pérdidas por absorción. Sólo para un caso se indica el efecto de la absorción en agua. Se estudia el efecto de las redistribuciones de 1311, 132 Cs y 47Ca por medio de un simulador de Alderson medido en la geometría de silla de 42 cm y en la geometría de camilla de exploración de 144/45 cm . También se cita un ejemplo de los estudios clínicos realizados sobre pacientes con 85Sr. La experiencia obtenida con la geometría de cam illa de exploración muestra que es posible evaluar la actividad del organismo entero con una exactitud de más del 15 por ciento, independientemente de su distribución en el cuerpo. La memoria sugiere geometrías. de exploración mejoradas en las que el paciente está tumbado en un arco, y otras en las que dos cristales se desplazan en órbitas curvas.

1. INTRODUCTION ■

In several applications of whole-body counting techniques a rather ac­ curate determination of the whole-body retention of radioactive tracers is desirable. For instance, in the case of human beings having a contami­ nation resulting from the occupational use of radioactivity or the adminis­ tration of small amounts of activity for diagnostic purpose, one often has to measure the whole-body retention of nuclides concentrated in only limited parts of the body. Moreover, the distribution of the tracer may also vary considerably during the period of a study. Intravenously injected nuclides, for example, are often initially rather uniformly distributed in the circu­ latory system. Many nuclides and labelled compounds rather quickly ac­ cumulate in particular organs of the body. In order to determine the absolute body activity, regardless of its distri­ bution in the body, it is desirable to have a detector arrangement with a counting efficiency that is independent of the site of the nuclide in the body. If this is not the case, the whole-body counter measures an apparent re­ tention, which must be corrected at frequent intervals for the varying counting efficiency.

2. THE CHAIR GEOMETRY

Several laboratories have reported results obtained by means of whole- body counting techniques with the "tilting chair”, which was originally de- A TWO-CRYSTAL SCANNING-BED COUNTER 57 veloped by M A R IN E L L I [1] and M IL L E R [2] in 1956. This geom etry was found to be useful for the measurement of approximately uniformly distri­ buted nuclides, such as potassium and C s 137 [3] . However, the method has proved to be less useful for non-uniformly distributed nuclides, as re­ ported by CEDERQUIST and LIDÉN [4], CEDERQUIST [5] and COHN et al. [б] and other authors. F ig.l shows the 42-cm chair geometry used in our laboratory. The Nal(Tl) crystal used has a diameter of 12.5 cm and athick- ness of 1 0 cm.

Fig. 1

The 42-cm chair geometry with a 12. 5-cm diam. by 10-cm thick N al(Tl) crystal in the iron room

3. THE SCANNING GEOMETRY

Whole-body counters for clinical use and for contamination examinations should utilize a detector arrangement that is able to smooth out the effects of re-distributions and inhomogeneities in the body. The geometry should also be comfortable for the patient. In our laboratory a geometry has been developed in which the patient lies on a stretcher (F ig . 2). One crystal is located above and one below the patient. The crystals (facing each other) can be moved continuously and simultaneously in a direction parallel to the patient from a position over the feet to a position over the head or vice versa. During a scan the pulses are fed into a gamma-ray spectrometer. The sum of the number of pulses in the photo peak from both .crystals is used as a measure of the whole-body activity. The distance between the detectors and the patient has been chosen after consideration of both the counting efficiency and the possibility of smoothing out the effects of inhomogeneities and re-distributions. As a compromise, it was decided to use a distance of 45 cm between the bed and the face of the upper crystal and 70 cm between the faces of the crystals. Thus the 58 Y . NAVERSTEN

The 144/45-scanning-bed geometry with two 12.5-cm diam. by 10-cm thick Nal(Tl) crystals. The crystals are located on each side of the patient. The distance between the faces of the crystals is 70 cm. detector faces fo r most subjects are at about equal distances from the body surface. The size of the counting room only allowed a scanning length of 144 cm. We refer to this counting technique as the 144/45 cm two-crystal, scanning-bed geometry*. . In the initial period of the development of the counting technique we used only one crystal. In this case two scans had to be made; one with the patient lying in an AP-position (anterior-posterior) and one with the patient in a PA-position. The sum of the number of pulses from both scans were then used as a measure of the whole-body activity. This technique gave results that were very similar to those of the two-crystal scanning-bed technique. Similar whole-body counting arrangements have also been reported by ROESCH and P A L M E R [7] and by E T O , W A T A N A B E , T A N A K A and H IR A M O TO [ 8 ] .

4. SOME DETECTOR CHARACTERISTICS

Most experience in the scanning geometry has been gained by using 12.5-cm diam. by 10-cm thick Nal crystals. The results here reported were obtained by counting with crystals of those dimensions. With these crystals it was noticed that for high energy gamma-ray photons from a point source the counting response in the photo peak was almost isotropic. Fig. 3 shows the relative counting efficiency for different angles between the axis of the crystal and the direction to the point source, obtained by measure­ ment of small sources of Co 57 and C s 137 about 100 cm from the centre of

* Another detector support in the counting room No. 2 made it possible to use a scanning length of 1 8 5 c m . A TWO-CRYSTAL SCANNING-BED COUNTER 59

F ig -3 '

The counting efficiency in the photo peak of 122-keV (Co57) and 662-keV (Cs137) y-ray photons at various angles between the axis of the crystal and the direction from the centre of the crystal face to the source. Normalization is made to the counting efficiency at 0°.

For Cs137 the efficiency obtained varied from 94% to 102%; for Co 57 from 82% to 114%. . L

the crystal face. For Csi37 the efficiency obtained varied from 94% to 102%, fo r Co57 from 82% to 114%. The lowest efficiency was obtained at 90° and the highest at about 45°. . F o r C s i37 the pulse amplitude in the full-energy peak of the 662-keV gam m a-ray photons on the average was altered less than 1%. The spectral resolution for the same peak was changed less than 1 0 %, indicating that the varying efficien cy is, to a rather sm all extent, due to a change in the shape of the spectrum. The efficiency is mainly governed by the effective surface exposed to the source. Results such as those shown in Fig. 3 for the 122-keV and the 662-keV gamma-ray photons justly permit us to regard the detector as an ideal point in the mathematical evaluation of the scanning geometry.

5. DEFINITIONS AND THE GENERAL FORMULA

Let us suppose there is a three-dimensional orthogonal co-ordinate system (x', y1, z'), see Fig. 4, and in this system the detector orbit is de­ fined by x' = h, y' = 0 , and -a < .z' S + a .■ . A finite orthogonal radioactive volume in the form of a parallelepiped . is placed on the y' z'-plane. The length, height and breadth of the volume are labelled, respectively, L, H and B. The co-ordinates of the corner closest to origin are (0, b, c). The volume is on all sides (to infinity) sur­ rounded by a non-radioactive absorber of the same material. A non­ radioactive absorber of height d is also placed directly on the radioactive volume. The same attenuation is assumed for both the radioactive and the non-radioactive volumes. Since the counting technique involves the use of pulses in the full-energy peak, a total linear attenuation coefficient ц has to be introduced in the calculation. If a small volume dx, dy, dz is at (x1, y ',z ') = (x, y+b, z + c)in the radioactive volume and if (h, 0 , u) is the position 60 Y . NAVERSTEN

The co-ordinate system with one detector orbit and a radioactive volume in the form of a parallelepiped

of the detector at a certain moment, then the distance between the small volum e and the d etector is

1 = [(h -x ) 2 + (y + b ) 2 + (z + c-u )2]*.

If the detector moves at a constant speed, the number of counts N during one scan of one detector during, the time t is '

, H + d-x exp (-Ml N = const X — X ------p ------dV du. (1)

V u

6 . GEOMETRICAL EFFICIENCY ,

, The geometrical efficiency in this report is defined as the counting .efficiency of an ideal point source, where no absorption between the source and detector is considered. In this case we get

h du N = const X -^X i = const X — ( 2 a) h2 + (c-u ) 2 ah

where a is half the scanning length, u describes the position of one detector in its scanning orbit, and с is the position of the source on a line parallel to the scanning orbit, see Fig. 5. The sum of the term s in brackets is easily A TWO-CRYSTAL SCANNING-BED COUNTER 61

F ig . 5

Schematic drawing showing the geometry of a point source in relation to two scanning detectors

identified as the angle under which the scanning orbit is seen from the source and Eq.(2a) can be replaced by

■ N = constX x j'd q ) = const X (ф2 -ф2). (2b)

The geometrical counting efficiency of point sources for a two-crystal scanning-bed geometry can be calculated from Eq.(2a) or Eq.(2b). F ig . 6 shows the geometrical iso-response curves for this geometry in the plane of the scanning orbits. The position of a human subject in the geometry we have chosen is indicated. It can be seen that over a length of more than 80 cm, where the trunk and part of the legs of an adult human being is normally situated, the geometrical efficiency does not deviate more than 1 0 % from the 1 0 0 % value in.the centre of the geometry.

Half scan

The geometrical iso-response curves in the plane of the detector orbits

Fig. 7 shows the geometrical iso-response curves in the central plane perpendicular to the detector orbits. Relatively speaking almost identical curve shapes (within a few per cent) are obtained in any parallel plane over 80% of the scanning length. However, in planes near the ends the response is much more homogeneous. In the plane shown in Fig. 7 the geometrical 62 Y . NA VERS TEN

' O ’uf "¿6"* 30 40 50 cm - i------1------:------1______i______i 1 Half scan = a (= 72 cm)

Fig. 7

The geometrical iso-response curves in the central plane perpendicular to the detector orbits efficiency for a tall man varies from 1 0 0 % to 130% in the sagittal direction and from 100% to 75% in the lateral direction. It is estimated that the geor metrical efficiency in an adequate air volume normally occupied by ahúman being can va ry by about a factor of three.

COUNTING EFFICIENCY WITHIN ABSORBING MATERIAL

If an absorber of thickness d is introduced between the point source and the detector, Eq. (2b) should be changed to

t Г d N = const X — X exp(-/u------)dcp. (3) ah f cos ш

F igs. 8 and 9 show the results of a numerical evaluation of Eq. (3). Fig. 8 shows the longitudinal and Fig. 9 the sagittal variation of the counting ef­ ficiency of a point source of Cs 137 in w ater (ц. = 0.086 cm -1) along the two

Half scan efficiency

1.0 ­

0.8 ­

777777777727777777777777777777 Without absorber 0.6 i— .!üp—_— QÁ- With absorber 'yiijiim Am m nii/i/m i/iiiiii/ii/îhfjl. ( ,37Cs in H20 ) 0.2

!%ee(= 25cm) 0 Ü ______

Fig.

The longitudinal variation of the counting efficiency of a C s 137 point source

displaced in the middle of a 2 0 -cm thick water phantom A TWO-CRYSTAL SCANNING-BED COUNTER 63

central lines of "the two-crystal scanning-bed geometry", normalized to the efficiency of the central point. The relative geometrical efficiency is also indicated. From F ig. 8 it can be seen that the variation in the longitudinal direction is the same, except near the ends, whether the source is "measured" with­ out an absorber or in the middle of a 2 0 -cm thick water phantom. A vertical displacement of the Csi37 source in water results in a more pronounced change of the efficiency (Fig. 9). For comparison the relative change of the geometrical efficiency is also shown. Without an absorber the efficiency interval is from 1.00 to 1.13. Over a 20-cm thick water phantom the interval is from 1.00 to 2.45.

Relative efficiency

■ F ig - 9 ■

The sagittal v a r ia tio n of the counting efficiency oi a CslJ1 point source

. displaced across a 2 0 -cm thick water phantom in the middle of the scanning length

8 . PHANTOM STUDIES

In order to get a m ore realistic idea of how much the counting efficiency varies with activity in different parts of the body a direct approach to the problem was made by means of phantom studies. An Alderson phantom with a human skeleton and sections simulating various organs of the body were filled with dilute radioactive solutions. Three nuclides, I131, Cs 132 and Ca47, with principal 7 - ray photon energies of 0.36, 0.67 and 1.31 MeV were studied in four parts of the mani­ kin - the whole body, the thyroid, the liver and the bladder. The manikin was counted in both the 42-cm chair geometry and the 144/45-cm scanning- bed geom etry. In the latter case the two-scan technique was used; one scan with the manikin lying in the AP-position and one scan with the manikin in the PA-position. Table I shows the response for the two different geometries, normal­ ized to the ease when all activity was uniformly distributed in the phantom (except in.the skeleton). The results in this table illustrate the improved capability of levelling out the effects of re-distributions in the body obtained with the scanning geometry as compared with the results obtained with the chair. In the chair the relative response for the different organs varied 64 Y . NAVERSTEN

TABLE I

THE RESPONSE OF 1131, Csl32 AND Ca47 IN THE THYROID, THE BLADDER AND THE LIVER normalized to the response obtained when the activity was uniformly distributed in the whole body, as obtained from measurements in both the 42-cm chair geometry and the 144/45-cm scanning-bed geometry.

The statistical accuracy is given as 2 SD.

N u clid e G e o m e try W h o le-b o d y T h yroid B ladd er L iver

1131 C h air 1 . 0 0 ± 0 . 0 2 1 .7 3 ± 0 . 0 7 0. 74 ±0.03 . 1. 84 ±0. 07

( 0 . 3 6 'M e V ) S c . bed 1 . 0 0 ± 0 . 0 2 0 . 9 6 ± 0 . 0 4 0 . 9 8 ± 0 . 0 4 1 . 0 7 ± 0 . 0 4

C s i3 z C h air 1 . 0 0 ± 0 . 0 2 1 . 6 8 ± 0 . 08 0. 80 ±0.04 1. 92 ± 0. 09

( 0 . 6 7 M eV ) S c . bed 1 . 0 0 ± 0 . 0 2 0 . 9 9 ± 0 . 0 5 1. 05 ±0. 05 1 . 1 0 ± 0 . 0 5

C a « C h air 1 . 0 0 ± 0 . 0 3 0 . 8 3 ± 0 . 0 4 2 . 0 0 ± 0 . 1 0

( 1 . 3 1 M eV ) S c . bed 1. 00 ±0. 07 - 1 . 1 1 ± 0 . 0 4 1 .1 3 ± 0 . 0 6

Efficiency ratio,

C h air 1. 69 ± 0. 03 - - - Scanning bed

(per crystal) between 0.8 and 2.0, i.e., by a factor of about 2.5. In the scanning geo­ metry the corresponding variation was from about 0. 95 to 1.15. From these results it is obvious that, for instance, in clinical appli­ cations, when nuclides after injection are initially rather homogeneously distributed in the body and later accumulate in particular organs, the measured retention becomes a matter for special attention. If the retention is referred to an initial measurement, when most of the activity is in the blood (e.g. Cr 61 in red blood cells and cobalt-nuclide-labelled vitamin B 1 2 , which may accumulate in the spleen and the liver) the chair geometry may give an apparent retention that is almost 100% higher than the real one. In such extreme cases the two-crystal scanning-bed geometry would give re­ sults that are in error by less than 15%.

9. RESULTS FROM CLINICAL STUDIES ■

Comparative results for directly measured retention and retention ob­ tained from excreta assays, for Sr85, Ca47, 1131 and even Co57 have been reported by CEDERQUIST and LIDÉN [4] and by CEDERQUIST (1964) [5]. Their results show a very good agreement between the retention measured with the scanning-bed geometry and the retention obtained from excreta measurements. Fig. 10 shows data obtained from comparable chair and scanning-bed measurements of two patients i.v. injected with about 0.5 juc Sr85. The A TWO-CRYSTAL SCANNING-BED COUNTER 65

7. Difference

Time after injection

F i g .10 . .

The percentage difference between the retention of Sr85, 1 to 10 d after injection : calculated from excreta measurements and measured by chair and scanning geometry in two patients. Square symbols denote results of the chair measurements and circular symbols results of the scanning measurements. The retention as calculated from excreta assays was used as a reference (zero). The bars indicate one standard deviation.

measured retention up to 1 0 days after the administration is ,compared with the retention obtained from excreta assay. During these days almost all retained activity is confined to the skeleton and thereafter negligible re­ distributions within the body will appear. On the ■ tenth day the chair measurements showed a retention of 1 0 % to 2 0 % lower than the real one. The retention obtained by the scanning technique did not deviate significantly from the retention calculated from excreta measurements.

10. CONCLUSIONS •

The results from measurements on a plastic manikin show that, regard­ less of where in the trunk of the human body the activity is situated, it is possible with the 144/45-cm two-crystal scanning-bed geometry to get re­ sults that are in error by less than 15%. The experience obtained from clinical studies with the same geometry show that the error in many practical cases is much smaller, even with the low-energy y-rays from Co 5 7 , . ■ ■ .

11. F U R T H E R IM P R O V E M E N T S ’ '

A still better evening out-of the effects of inhomogeneities would be ob­ tained with a longer scanning length and a larger distance between the scanning orbits. Gedmetries with scanning lengths up to 185 cm are under investigation in our laboratory, parallel with the evaluation of formulae similar to the E q .(l).j However, the counting efficiency is to a first ap- 66 Y . NAVERSTEN

Half scan

The geometrical iso-response curves from one scanning detector. . The thick lines indicate the stretcher and the suggested 2-metre arc support for the patient. proximation inversely proportional to the parameters a and h. An increase of these parameters would therefore not favour our aim of reducing the minimum detectable activity as much as possible. Another way to get a counting geometry that is less dependent on the body distribution of the activity is to use an arc instead of the orthodox stretch er. Such a geom etry is suggested in F ig . 11. A patient lying on a 2-metrè arc would have the geometrical iso-response curves of one scanning detector approximately parallel to the body. If absorption effects are also considered, optimal counting characteristics could be obtained by using an arc with a somewhat longer radius. Such a geometry, with measurements in both the prone and supine positions, would have about the same advantages as EVANS arc geometry [9], but it would have a higher counting efficiency; it would not take up so much space in the counting chamber; and it would be more comfortable for the patient than is the one-metre arc.

ACKNOWLEDGEMENT

I wish to express my sincere gratitude to Dr. Kurt Lidén, who sug­ gested this investigation, for his kind aid and advice.

■ REFERENCES

[1] MARINELLI, L.D. , Brit, J. Radiol. Suppl. 7 (1957) 38. [2] MILLER, C.E. , Rpt ANL-5596 (1956) 26. [3] MILLER, C.E. , MAY, H. A. and MARINELLI, L.D. , Proc. Symp. Health Physics in Nuclear Installations, 25-28 May 1959, OEEC and Danish Atomic Energy Commission, Ris0 (1959) 119. [4] CEDERQUIST, E.S. and LIDÉN, К. V. H. , Whole-body counting, IAEA, Vienna (1962) 487. [5] CEDERQUIST, E ., Acta Radiol. Suppl. (1964).

[6 ] COHN, S .H ., LIPPINCOTT, S. W. , GUSMANO, E. A. and ROBERTSON, J. S. , Rad. Res. _HS(1963) 104. [7] ROESCH, W.C. and PALMER, H.E. (Private communication). '

[8 ] ETO, H ., WATANABE, H. , TANAKA, E. and HIRAMOTO, T ., Whole-body counting, IAEA, Vienna (1962) 211.

[9] EVANS, R.d ’, Amer. J. Roentgenol. 37 (1937) 368. , A NEW TECHNIQUE FOR DETERMINING THE DISTRIBUTION OF RADIUM AND THORIUM IN LIVING PERSONS

C.E. MILLER* ARGONNE NATIONAL LABORATORY, HEALTH DIVISION ARGONNE, ILLINOIS, UNITED STATES OF AMERICA

Abstract — Résumé — Аннотация — Resumen

A NEW TECHNIQUE FOR DETERMINING THE DISTRIBUTION OF RADIUM AND THORIUM IN LIVING PERSONS. Whole-body counters have been traditionally used to measure radioactivity within persons without reference to internal distribution. In the case of bone-seeking radioisotopes, such as radium, it is important to have some knowledge of at least the gross distribution within the skeleton in order to study the nature of the dose response. The techniques described here were developed to detect substantial non-uniformity in distribution of the radium in patients who cannot or will not lie still for more than one hour. A newly-designed right-cylindrical crystal 15 cm diam. and 20 cm long is placed 30 cm above the surface of a rigid bed with the axis of the crystal horizontal and at right angles to the supine patient's vertebral column. This log-shaped crystal is used instead, of the typical vertical squat cylinder in order that the surface areas and cross-sections of the crystal be of the same shape when viewed from any point in the body. The gamma-ray spectra from the supine patient and from point radioactive sources in the centre of Presdwood phantoms are obtained with the crystal located at x number of uniformly-spaced positions along the bed. The counting rates obtained for selected photopeaks from the spectra of the patient and the phantom are fed into an electronic computer, which is programmed to calculate a series of point sources that would give the same gamma-ray profile. i The results with a number of radium patients have demonstrated that, while the distribution is frequently essentially uniform, some persons contain a concentration two or three times higher than the mean in the lower legs, the pelvis, or in the skull. The sums of the calculated sources agree within a few per cent with the total radium body contents measured with other whole-body counting techniques. , Measurements made o f two patients who received thorotrast about eighteen years ago and of phantoms containing thorotrast and RdTh have also demonstrated the feasibility of measuring the non-uniformity of

distribution of Tl 208 (ThC"), of Ac 228 (MsTh2), and possibly of Pb212 along the axis of the body.. •

DÉTERMINATION DE LA RÉPARTITION DU RADIUM ET DU'THORIUM CHEZ LES PERSONNES VIVANTES: UNE MÉTHODE NOUVELLE. En règle générale, on se sert d'anthropogammamètres pour doser l'activité du corps humain, sans se préoccuper de sa répartition. Dans le cas des radioisotopes ostéotropes* comme le radium, il importe d‘avoir une idée au moins approximative de leur répartition dans le squelette pour étudier la nature de la réponse à la dose. Les techniques décrites dans le mémoire ont pour objet de déceler toute inhomogénéité importante dans la répartition du radium chez des malades qui ne peuvent pas ou ne veulent pas rester allongés et immobilisés pendant plus d'une heure. • Un cristal de conception nouvelle, en forme de cylindre droit, mesurant 15 cm de diamètre sur 20 cm de long, est situé à 30 cm au-dessus de la surface d'un lit rigide, l'axe du cristal étant horizontal et orthogonal par rapport à la colonne vertébrale du malade en supination. Ce cristal horizontal allongé est utilisé de pré­ férence au cylindre vertical court de type courant, de manière que la surface et les sections du cristal aient la même forme, quel que soit le point du corps d‘oti on les observe. Le cristal étant placé à un nombre x de positions régulièrement espacées le long du lit,, on détermine les spectres des rayons gamma émis par le malade en supination et par des sources radioactives ponctuelles placées au centre dé fantômes en bois artificiel (Presdwood). Les taux de comptage obtenus pour des pics photoélectriques déterminés provenant des spectres du malade et du fantôme sont transmis à une calculatrice électronique, laquelle est programmée de manière à calculer une série de sources ponctuelles qui fourniraient le même profil de rayons gamma.

* Work performed under the auspices of the USAEC.

67 68 C .E . MILLER

Les résultats obtenus chez un-certain nombre de malades traités au radium montrent que, si la répartition du radioisotope est souvent essentiellement uniforme, chez certaines personnes sa concentration est deux ou trois fois supérieure à la moyenne dans la partie inférieure des jambes, dans le pelvis ou dans le crâne. Les sommes des sources calculées'concordent, à quelques pour-cents près, avec'les mesures de la charge corporelle totale de radium obtenues au moyen d’autres techniques d’anthropogammamétrie. , Les mesures faites sur deux malades qui avaient reçu.du thorotrast il y a environ dix-huit ans ainsi que sur des fantômes contenant dli thorotrast et RdTh ont également mis en évidence la possibilité de mesurer l'inhomogénéité de la répartition de’ 208T l ( T h C " ) , 228Ac(MsTh2), et aussi sans doute de zl2 Pb, le long de l 'a x e du corps. ‘ '

НОВЫЙ МЕТОД ОПРЕДЕЛЕНИЯ РАСПРЕДЕЛЕНИЯ РАДИЯ И ТОРИЯ У ПАЦИЕНТОВ. Обычно для измерения радиоактивности у человека без учета распределения внутри организма применялись счетчики для измерения активности всего организма. Когда речь идет о таких оседающих в костях радиоизотопах, как радий, важно иметь некоторые сведения по крайней мере об общем распределении их в скелете для того, чтобы изучить отношение дозы и эффекта. Описываемая методика была разработана для определения существенной неоднородности рас­ пределения радия у пациентов, которые не могут или не хотят лежать спокойно больше .часа. Заново сконструированный прямой цилиндрический кристалл диаметром 15.и длиною 20 см поместили по горизонтальной оси на расстоянии 30 см над поверхностью специально устроенной жесткой кровати под прямым углом к позвоночнику лежащего пациента. Эта вытянутая форма кристалла была применена вместо обычного вертикально расположенного короткого цилиндра для того,-чтобы площадь поверхности и поперечное сечение кристалла имели одинаковую форму по отношению к любой точке тела. . Спектры гамма-лучей, исходящих от позвоночника пациента и от радиоактивных источ­ ников в центре фантома из прессованной древесины, были получены при помощи кристалла, помещаемого в х-числе положений с одинаковыми интервалами вдоль кровати. Скорости счета, полученные для отобранных фотопиков из спектров пациента и фантома, были введены в электронносчетное устройство, запрограммированное для подсчета серии точечных источни­ ков, что дает тот же самый профиль гамма-лучей. . : Результаты, полученные при исследовании ряда пациентов, свидетельствуют о том. что хотя распределение чаще всего в основном однородно, у некоторых лиц концентрация в два или три раза выше средней в ногах, тазу или в черепе. Итоги вычисленных источников соответ­ ствуют в пределах'нескольких процентов общему содержанию радия в организме, подсчитан­ ному на основании других методик измерения активности всего организма. Измерения, проведенные у двух пациентов, получавших торотраст более девятнадцати лет назад, и на фантомах, содержащих торотраст и радий-торий, также свидетельствовали

о возможности измерения неоднородного распределения Tl 20á(ThC"), Ac 22á(M sTh2) и, возможно,

РЬ 212 вдоль оси тела. ; ' ’ ’

NUEVA TÉCNICA-PARA DETERMINAR LA DISTRIBUCIÓN DEL RADIO Y DEL TORIO EN PACIENTES. Los antropogámmámetros se vienen usando tradicionalmente para medir la radiactividad de las personas sin atender-a la distribución interna. En el caso de los radioisótopos osteófilos, tales como el radio, interesa conocer por lo menos su distribución aproximada dentro del esqueleto, a fin de poder estudiar la naturaleza de la respuesta a la ‘dosis. Las técnicas descritas en la memoria se'idearon’para poder descubrir cualquier heterogeneidad sustancial en la distribución del radio en pacientes que no pueden o no desean yacer inmóviles más de una hora. . • ' ■ Un cristal cilindrico recto, de diseño nuevo,' de 15 cm de diámetro por 20 cm 'de longitud,- se coloca a una altura de 30 cm sobre -la superficie de una camilla rígida, con el eje del cristal horizontal y perpendicular á la dirección de la columna vertebral del paciente, acostado en posición supina. Se usa este cristal en lugar dél cilindro vertical chato tfpico, con la intención de que las áréas superficiales y las secciones transversales del cristal, observadas desde cualquier punto del cuerpo del paciente, tengan la misma forma. ' Se obtienen los espectros de rayos gamma del paciente en posición supina y de las fuentes radiactivas puntiformes ubicadas en el centro de maniquíes de madera prensada,' con el cristal situado en un número x de posicionés uniformemente espaciadas a lo largo de la cam illa. Los recuentos obtenidos en correspondencia con determinados picos fotoeléctricos de los espectros del paciente y del maniquí se-pasan a una calculadora electrónica cuyo programa se establece'a fin de'calcular una serie de fuentes puntiformes que darían el mismo perfil de rayos gamma. ; ' • • Los resultados obtenidos con cierto número de pacientes tratados con radio han demostrado que, si bien là distribución es con frecuencia esencialmente uniforme, en algunas personas la concentración en los extremos RADIUM AND THORIUM IN. LIVING PERSONS 69 de las piernas, la pelvis, o el cráneo es dos o tres veces superior al valor medio. La suma de las fuentes calculadas concuerda, con error inferior a algunas unidades por ciento, con el contenido total de radio, medido por otras técnicas antropogammamétricas. - • , Las determinaciones efectuadas en dos pacientes a los cuales se les había suministrado una solución de torotrasto unos dieciocho años atrás y en maniquíes que contienen torotrasto y RdTh han demonstrado que también es posible evaluar la heterogeneidad en la distribución del 208Tl(ThC"), del 228Ac (MsTh2) y, quizás, también del 2i2Pb a lo largo del eje corporal.

INTRODUCTION

A whole-body counter that is used for the measurement of radium in man should provide as much information as possible regarding the distri­ bution of radium within the body in order to correlate body burden with clini­ cal findings. The standard tilting^ chair technique [1] provides a single measure of total body burden, of the gamma-emitting isotope under study, m is measurement can be used to calculate only the average dose rate ab­ sorbed by the tissues where the isotope is deposited. In the case of bone- seeking radioisotopes such as radium, it would be of value tó have additional information concerning uniformity or non-uniformity of distribution of the isotope in various major regions of the body. As an extension of the single-position, single-crystal technique, a single-crystal, multiple-position procedure was devised to reconstruct the gamma-ray profile along the long axis of the body without the use of a col­ limated crystal, which would require heavy shielding. The gamma-ray profile was simulated by computing the values for seven equally spaced point sources in the body that would give the same readings as those observed when the crystal is successively moved to sevenpositions along the body. - This procedure assumed that for each arbitrary segment of the body the radium content is the same as if it were concentrated in a "point" at the middle of the segment. In order to calculate, from .the .observed measure­ ments, the magnitude of these point sources in the body, a series of standard sources was measured in Presdwood phantoms, and the two sets of values were entered into a set of simultaneous equations. Since the sum of the magnitudes of radium in each of the seven segments of the body was equal to the patient1 s total content of radium, this procedure provided a whole- body count as w ell as an. approximate description of the extent of uniformity of deposition of radium in the skeleton. ■. , • This technique was developed in order to.determine the gross distri­ bution of radium.in patients who are unwilling or who are physically unable to be measured for more than one hour. Since most of the individuals of clinical value contain only 0.03 /кс to Í.0 дс gamma-ray emitting radium. В and С distributed throughout 7000 g of bone or 3500 cm 3 of tissue, many hours would be required to make a complete scan of their skeletons by con­ ventional medical scintillation scanning equipment. The multiple-position technique described here will identify those patients that have a non-uniform distribution, and these cases can be considered separately for evaluation of differential dose-response within various regions of the body. A conventional scintillation scan can then be made of-selected areas of high isotope con­ centration in these individuals. 70 С .E. MILLER

CRYSTAL-PATIENT GEOMETRY

A newly-designed, right-cylindrical Nal(Tl) crystal, 15 cm in diameter by 20-cm long, is suspended 30 cm above the surface of a rigid bed with the axis of the crystal horizontal and transversely placed with respect to the long axis of the patient. This log-shaped crystal is used instead of the usual squat cylinder with vertical axis in order that the surface area and cross-section of the crystal be essentially the same when viewed from any point in the body. Thus, the shape of the spectrum is the same whether the activity is located directly beneath the crystal (Fig. 1, point A) or at a con­ siderable distance from it (Fig. 1, point B). This situation does not obtain when the squat right cylinder is placed above the supine patient because the portion of the body directly under the crystal is viewed principally by the face of the crystal while the more distant parts of the body are viewed both by the face and the side. The equivalence of spectra produced by the log- crystal is of particular importance when spectrum stripping is necessary in order to determine the independent distribution of two or more isotopes in the body, such as members of the thorium decay chain.

I 2 3 4 5 6 7 crystol positions 73.5 49 24.5 0 24.5 49 73.5 cm from centre

Cross-section geometry of log-shaped scintillation crystal. The shape and cross-sectional area are essentially the same when viewed from any point in the outstretched body.

The log-crysta l is supported on a carriage that can be positioned at any location along the length of the patient. For the measurements discussed in this paper, the axis of the crystal was accurately placed at each of seven uniformly spaced positions 24.5 cm apart. The patient was placed on his back on a Lucite slab so that his longitudinal mid-point was always directly under crystal position 4 (Fig. 2). The patient's arms rested on the surface of the table alongside his body. Gamma-ray spectra were measured with the crystal placed over the patient at each of seven positions starting with number 1 at the head. Each measurement lasted for so long that each se­ lected energy band usually contained at least 1 0 0 0 0 counts or had a standard deviation of ± 1 %. ,

CALIBRATION

Reference spectra for analysing the data from patients were obtained from measurements of standard radium sources in a Presdwood phantom. RADIUM AND THORIUM IN LIVING PERSONS 71

I 2 3 4 5 6 7 crystal positions 73.5 49 24.5 . O' 24.5 49 73.5 cm from centre

Diagram of the single-crystal, multiple-position arrangement used in the measurement of the outstretched subject.

6 7 crystol positions 49 73.5 cm from centre ж . ж

f + - , + 1 + - + \ - + j ••А.У -A.* '■ + ^— ~ T ~ 30 cm 1 phantom thickness in centimetres '

Fig- 3

Diagram of the arrangement of Presdwood segments used as a phantom in calibration of the single-crystal, multiple-position system.

Sheets of Presdwood 30 cmX 30 cmX 0.6-cm thick were placed in seven stacks of varying heights to approximate to the body profile obtained by aver­ aging the anterior-posterior diameters of a series of patients (Fig.3). A 1 . 0 цс radium source was placed in the centre of the first phantom segment directly under crystal position 1 , and the gamma-ray spectra were recorded with the crystal placed successively at each of the seven positions. This series of measurements was repeated with the radium source placed in turn in the centre of the second through to fifth phantom segments and with the source placed 2-cm deep in segments 6 and 7 in order to approximate the location of the long bones in the legs. The thickness of the phantom at any position need not be exactly the same as the thickness of the patient for the following reasons. With the crystal 30 cm above the surface of the Lucite slab, the shapes of the spectra and the counting rates are almost identical when the radium source is in the centre of a phantom that is 8 cm, or 15 cm, or 22 cm thick. Since the distance from the crystal to the base of the phantom remains fixed at 30 cm, the source is closer to the crystal in the thicker phantoms, but the increased count rate resulting from .the reduced crystal-to-source distance is counteracted by the higher absorption in the increased interposed mass. If the crystal were further from the surface of the bed, the counting rate would vary as a function of phantom thickness. The 30-cm height was selected because it is the minimum table-to-crystal distance that accommodates the majority of the patients. 72 C.E. MILLER

ANALYSIS OF DATA

The magnitude of the simulated seven-point sources in the patient may be computed by entering the human and phantom measurements into seven simultaneous equations. The counting rates in five selected energy bands from the appropriate spectra were substituted, one band at a time, in the following equations:

С i = R iX u + R2X 12 + R3X 13 + R4 X 14 + . . . + R 7X 17 C2, = R1X 2 1 + R 2X 2 2 + R 3X23+... + r 7x 27 ; : ’ (D

Cy = R2X 72 + R 2X 72 + . . . +R,7X 77 w here: , C1; C2, . . . are the actual counting rates obtained from the pa.tient at the seven crystal positions denoted by the subscripts; Х ц , X i2, . . . are the counting rates obtained from the phantom measure­ ments, the first digit of the subscript' denoting the position of the crystal and the second the location of the source; Ri, R 2, • • • are the quantities of radium at each of the seven positions that would yield the same radioactive profile as that observed in the patient. The values for R are calculated by solving the seven simultaneous equations. The equations were solved by an IBM-704 electronic computer for which a general programme for matrix inversion and solution of linear equations was available. A punch card input system was used to enter the constant vectors (Ci, C2, ...) and the m atrix coefficients (Х ц , X i^ ...). The printed output listed the solution vectors (Ri, R2, . • . ) for each energy band and each patient. ' In order to provide an internal test of the procedure, to test the phantom spectra against the human spectra, and to determine the statistical accuracy of the various point sources, the analysis was performed with the counting rates in five different energy bands. These energy bands were: 75-275, 275-375, 525-675, 775-1275 and 1575-2575 keV. The first two bands mainly include gamma rays that are scattered in the body while the higher energy bands represent unscattered primary gamma rays. The magnitude of the seven sources determined on the basis of each of the first two bands would be considerably different from those found with the higher energy bands if- the radium were in fact located at a different effective distance in the patient than in the phantom. Since the magnitude of the seven sources found with five energy bands agreed quite well with ea:ch other, there is justification for assuming that the situation in the patient can be simulated by a point source in the phantom. . The detailed values for Case No'. 03-571, a radium dial painter, are shown as a- typical example of the results obtained with a patient whose radium distribution is fairly uniform (Table I). The data for a patient, who received injections of radium for medical purposes, Case No. 03-206, are also presented since they demonstrate fairly marked non-uniformity of radium deposition. TABLE I QUANTITY OF RADIUM CALCULATED FOR SEVEN SEGMENTS OF THE BODY . FOR TWO RADIUM-CONTAINING PATIENTS

Crystal position

Energy band 1 2 3 4 5 6 ' 7 T o ta l PERSONS LIVING IN THORIUM AND RADIUM . (k e V )

Case No. 03-571

. 7 5 - 2 7 5 . 0 5 0 4 .0 4 1 1 . 0 2 4 7 . 0 4 8 5 . 0 3 7 6 . 0 4 8 4 . 0 4 4 8 . 2 9 5 4 4

2 7 5 - 3 7 5 . 042 0 . 0 3 8 1 . 0 2 4 5 . .0 4 7 3 . 0 3 6 6 . 0 5 0 7 . 0 3 3 0 . 2 7 2 3

5 2 5 - 6 7 5 .0 5 2 7 ' . 0 3 8 9 . 0 2 4 8 . 0 4 2 9 . 0 3 8 2 . 0 4 8 9 . 0 3 9 9 . 2 8 6 4

7 7 5 - 1 2 7 5 .0 5 4 1 ‘ . 0 4 0 1 . 0 2 6 3 . 0 4 0 0 . 0 3 8 6 . Ó487 . 0 4 4 1 . 2 9 1 9

1 5 7 5 - 2 5 7 5 . 0 5 0 4 . 0 4 1 1 . 0 2 4 7 . 0 4 8 5 . 0 3 7 6 . 0 4 8 4 . 0 4 4 8 . 2 9 5 4

M ean , 0 4 9 9 . 0 3 9 9 . 0 2 5 0 . 0 4 5 4 . 0 3 7 7 . 0 4 9 0 .'0 4 1 3 . 2 8 8 3

S. E. of mean ± . 0 0 2 1 ± . 0 0 0 6 ,± . 0 0 0 3 ± . 0 0 1 7 l . 0 0 0 3 ± .0 0 0 4 ± . 0 0 2 3 ± . 0 0 4 3

Case No. 03-206

7 5 - 2 7 5 . .1 3 0 0 . 0 7 1 8 . 0 2 7 6 . 0 6 1 9 . 1 5 3 7 , . 2 8 7 1 . 1 3 6 9 . 8 6 8 8

2 7 5 - 3 7 5 .1 5 4 2 . 058 1 . 0 3 2 1 . 0 5 2 9 . 1 4 5 8 . 2 8 5 4 . 1 4 6 4 . 8 7 4 9

5 2 5 - 6 7 5 . 1 6 5 4 . 0 6 9 7 . 0 3 9 4 . 0 5 3 5 . 1 5 9 0 . 3 0 1 3 ^ . 1 6 6 8 . 9 5 5 2

, 7 5 5 - 1 2 7 5 ’ .1 6 2 3 . 0 7 7 5 . 0 3 2 1 . 0 5 6 7 . 1 7 0 5 , 2 9 8 2 . 1 8 0 5 . 9 7 8 0

1 5 7 5 - 2 5 7 5 • .1 6 4 7 .0 7 0 0 . 0 3 6 7 . 0 4 7 7 . 1 7 1 9 , 3 1 1 0 .1 5 3 5 . 9 5 5 4

M ean . 15 5 3 . 0 6 9 4 . 0 3 3 6 . . 0 W 5 . 160 1 . 2 9 6 6 .1 5 6 8 . 9 2 6 4

S. E. of mean 1 . 0 0 6 7 ± . 0 0 3 2 ± . 0 0 2 1 ± . 0 0 2 3 ± . 0 0 5 0 ± .0 0 4 7 ± .0 0 7 7 i . 0 2 2 7 74 С . E. MILLER

RESULTS

Radium

Thirty patients were measured with both the multiple-position technique and with the standard tilting-chair technique. Summary data for seven of these patients are given in Table II. The right-hand columns give the total body radium estimated by the multiple-position and the tilting-chair methods. The first seven columns give the calculated radium values in microcuries for each seventh segment of the body, and the percentage value directly beneath. . The seven Calculated points, which vary in magnitude from 0.025 цс to 0.05 у с for patient 03-571, demonstrate a fairly uniform distribution of radium along the skeleton. In contrast, data from some of the patients show a very non-uniform distribution. For example, 03-423 and 03-206 have 27% and 32%, respectively, of their t o t a l body activity under crystal position 6 , while patient 03-115 has 32% under crystal position 4, and patient 03-110 has 33% under crystal position 1. The data obtained for this group of thirty patients demonstrate that from 5% to 32% of the total body radium can be found in any seventh segment of the body. The total-body measurements obtained by the seven-position technique and by the tilting chair are in very good agreement for the first four patients in Table П. However, the tilting-chair measurement is only 85% of the seven-position value for the remaining three patients. This difference can be explained, at least in the case of the last two patients (03-110, 03-402), on the basis of the distribution of radium in the body. Here, where 26% to 3 3 % of the radium is in the head, the crystal used with the tilting chair has a different geometry with respect to the head than to the remainder of the body, since the head faces the side of the crystal where the canning material is thickest (Fig. 4). , ■

Thorium

Two patients who were injected with thorium dioxide eighteen years ago were similarly measured and analysed with the seven-position technique. The system was calibrated for these measurements by counting a thorium dioxide source and a radiothorium source at each of the crystal positions in a phantom. By this technique it was possible to strip apart the observed spectra and to determine the individual distribution of both MsTh2 and ThC" in the body so that the concentration of M sTh2 in the liv er was separated from the ThC" seen in the long bones. With further analysis it may also be possible to strip apart the contributions from Pb212 and to establish a gross distribution for that isotope as well.

ACCURACY

The inter-relationships of the terms in the set of simultaneous equations is such that an e rro r in measurement at one of the crystal positions w ill result in a sizeable error in the calculated source at the same position and TABLE II

SUMMARY OF MEASUREMENTS OF SEVEN SELECTED PATIENTS CONTAINING RADIUM

Sum of T iltin g Tilting chair . Crystal position 7 positions c h a ir 7 positions PERSONS LIVING IN THORIUM AND•RADIUM P a tie n t

(u n it) 1 2 3 4 5 6 . ' 7

0 3 -5 7 ": (JJC) . 0 4 9 9 .0 3 9 9 . 0 2 5 0 .0 4 5 4 .0 3 7 7 .. 0 4 9 0 .0 4 1 3 . 2 8 8 3 . 2 9 6 1 . 0 2 8

(%) 1 7 . 3 1 3 . 8 8 . 8 1 5 . 8 1 3 . 1 1 7 . 0 1 4 . 3 1 0 0 . 0

0 3 - 1 3 9 ( ^ c ) . 0 2 7 1 .0 1 0 3 . 0 1 4 0 .0 1 6 5 . 0 3 0 9 . 0 2 3 5 . 0 1 8 9 . 1 4 1 2 . 1 4 5 9 1 . 0 3 3

№ 1 9 .2 7 . 3 9 . 9 1 1 . 7 2 1 . 9 1 6 . 7 1 3 . 4 1 0 0 . 0

0 3 - 2 0 6 (MC) . 1 5 5 3 .0 6 9 4 . 0 3 3 6 . 0 5 4 5 . 1 6 0 2 . 2 9 6 6 . 1 5 6 8 . 9 2 6 . 978 1 . 0 5 6

C io) 1 6 . 8 7 . 5 3 . 6 5 . 9 1 7 . 2 3 2 . 0 1 6 . 9 1 0 0 . 0

0 3 - 4 2 3 (PC) .0 1 8 1 .0 3 1 3 . 0 0 8 9 . 0 2 2 7 . 0 2 3 6 . 0 4 9 4 .0 2 9 1 " . .1 8 3 2 . 1 8 5 1 . 0 1

(%) 9 . 9 1 7 .1 4 .-9 1 2 . 4 1 2 . 9 2 7 . 0 1 5 . 9 1 0 0 . 0

0 3 - 1 1 5 0 « 0 . 0 4 7 6 .0 4 2 1 .0 3 4 9 . 0 8 5 9 . 0 2 3 3 . 0 1 4 2 . 0 2 0 3 . 2 6 8 2 . 2 2 8 . 8 5 4

(%) 17 -. 7 1 5 . 7 1 3 . 0 3 2 . 0 8 . 7 5 . 3 7 . 6 1 0 0 . 0

0 3 - 1 1 0 (Me) . 0 7 7 2 .0 3 1 7 . 0 2 0 2 . 0 4 3 6 .0 2 5 4 . 0 1 7 2 . 0 1 8 8 .2 3 4 3 . 1 9 5 6 . 83Ü

(% ) . 3 2 . 9 1 3 . 6 8 . 6 1 8 . 6 1 0 . 9 7 . 3 8 . 1 1 0 0 . 0

0 3 - 4 0 2 (MC) . 1 1 6 6 . 0 7 2 1 . 0 5 2 7 . 0 7 3 4 . 0 3 1 5 . 0 6 4 4 . 0 2 9 8 .4 4 0 6 . 3 7 . 84

(%) 2 6 . 5 1 6 . 4 1 2 . 0 1 6 . 7 7 . 2 1 4 . 6 6 . 8 1 0 0 . 0 76 C.E. MILLER

Diagram of the standard tilting-chair technique with the squat right cylinder in place. in smaller errors in the adjacent positions. For example, if the actual measurement at crystal position 4 is X% high, then the calculated value of the simulated source fo r that position w ill be 4X% high, and the sources cal­ culated for positions 3 and 5 will be 2X% low. The values calculated for the more distant crystal positions will not be materially affected by an error at position 4. Since the error of the calculated source is four times the error of the measurement of the patient, it is necessary to determine the counting rate at each position within a standard uncertainty of less than one-fourth of the desired uncertainty in the calculated source.. If this technique were expanded to include 13 positions involving 13 simultaneous equations, the measurements of the patient would ha:ve to be determined with much higher accuracy since any error in these measure­ ments would be multiplied by a correspondingly larger factor. Thus, this technique cannot be expanded to a 13-point or 2 0 -point analysis unless the activity at each point is measured with sufficient accuracy. For this reason it is not ordinarily possible to use more than seven points if the measurement time is to be restricted to one hour. The magnitudes of the seven calculated point sources would be more accurate if the spacing between the seven crystal positions were always one seventh of the patient's height. This condition would require a remeasure­ ment of the phantom calibration spectra for each particular crystal spacing used. In practice, however, the range of heights ordinarily encountered in adult patients does not lead to a displacement of more than 7 cm for crystal positions at the extreme ends of the body for persons who are quite short (158 cm) or quite tall (188 cm). The displacement for persons of inter­ mediate heights will of course be much less. RADIUM AND THORIUM IN LIVING PERSONS 77

SUMMARY

A single-crystal, multiple-position technique has been developed for measuring the body content of gamma-emitting isotopes in human subjects. The results obtained from measurement of thirty patients who contain radium that was administered over thirty years ago demonstrate that the radium is not always uniformly distributed within the skeleton. Some patients have as much as 33% of their total body radium in a relatively small volume of the skeleton, such as the skull. The relatively good agreement between whole-body measurements based on the multiple crystal-position and on the tilting-chair techniques demonstrates that the latter is fairly accurate for most patients. However, the data also suggest that higher accuracy will be obtained for radium measurements with the newer technique. In addition, this method provides some information about the distribution of radium in the skeleton so that those patients who have a grossly non-uniform distribu­ tion may be studied further by use of other techniques. This method may also be applied to the study of thorium deposition in the body.

ACKNOWLEDGEMENTS

This work is part ¡of a co-operative study of the long-term effects of radium deposition in man currently being conducted with Dr. R. J. Hasterlik, Argonne Cancer Research Hospital and Dr. A. J. Finkel, Argonne National Laboratory. I am grateful to Dr. Finkel for encouragement and support of the work tha!t led to the present paper. I wish to thank Mr. W. J. Snow, Applied Mathematics Division, Argonne National Laboratory, for adapting the computer programme to accommodate these data. Thanks are also ex­ tended to Mr. J.B.. Corcoran for technical assistance in collecting a portion of the data.

REFERENCE

[1] MILLER, C .E. , "An experimental evaluation of multiple-crystal arrays and single-crystal techniques”, Whole-Body Counting, IAEA, Vienna (1962) 81.

SHAPES OF SCINTILLATION SPECTRA

K. G. McNEILL AND V. K. MOHINDRA • ( ’ PHYSICS DEPARTMENT, UNIVERSITY OF TORONTO, CANADA 1

Abstract — Résumé — Аннотация — Resumen

SHAPES OF SCINTILLATION SPECTRA. A series of experiments have been carried out to find the effect on the shape of scintillation spectra of self-scattering in the source m aterial. This has application to the analysis of spectra obtained from persons internally contaminated with a mixture of unknown isotopes. Results with a uniform distribution of radioactivity in phantoms have previously been published [1]. The present paper describes experiments with point sources of gamma-rays. As criteria of spectral shape, the effective resolution of the photopeak and the valley-to-peak ratio of the photopeak are taken. As variables, we have considered the detector-to-source distance, the scatterer thickness, the depth of the source in the scatterer, and the gamma-ray energy. As examples of the results obtained, the valley-to-peak ratio for a 5-in Nal crystal varies by a factor of six as the depth of a source of 0.28 MeV gammas alters from 0 to 22 cm in a scatterer of fixed geometry, while for 1.1 MeV gammas the corresponding factor is two. There are related changes in effective resolution, though the factors are not so large. Graphs will be presented showing these changes and comparing them with the corresponding situations with uniformly distributed sources.

FORMES DES .SPECTRES DE SCINTILLATION. Les auteurs ont fait une série d’expériences pour déterminer les effets que Г autodiffusion, dans la matière qui constitue la source, peut avoir sur la forme des spectres de scintillation. La question se pose pour l'analyse des spectres observés chez des personnes présentant une con­ tamination interne due à un mélange de radioisotopes non identifiés. Les résultats obtenus pour une répartition uniforme de la radioactivité dans des fantômes ont déjà fait l'objet d'une communication [1]. Le mémoire décrit des expériences faites au moyen de sources ponctuelles de rayons gamma. Comme caractéristiques de la forme du spectre, les auteurs ont pris la résolution effective du pic photoélectrique et le rapport entre le creux et la crête du pic photoélectrique. Ils ont considéré comme variables la distance entre le détecteur et la source, l’épaisseur du diffuseur, la profondeur de la source dans le diffuseur et l’énergie des rayons gamma. Les auteurs parviennent notamment aux conclusions suivantes: pour un cristal de Nal de 12,5 cm , le rapport entre le creux et la crête varie d'un facteur de 6 lorsque la profondeur à laquelle une source de rayons gamma de 0,28 MeV est située dans un diffuseur de géométrie fixe passe de 0 à 22 cm ; pour des gamma de 1,1 MeV, le facteur correspondant est de 2. La résolution effective varie également, mais les facteurs ne sont pas aussi élevés. Les auteurs présenteront des graphiques qui représentent ces variations et permettent de les comparer aux résultats correspondants obtonus pour des sources uniformément réparties.

ВИДЫ СПЕКТРОВ СЦИНТИЛЛЯЦИЙ. Проведена серия экспериментов для определения влияния процесса саморассеяния в материале источника на форму сцинтилляционного спектра. Данные можно использовать при анализе спектров, полученных у лиц, подвергшихся внутрен­ нему заражению смесью неизвестных изотопов. Ранее были опубликованы данные с однород­ ным распределением радиоактивности в фантомах (11. В данной работе дается описание экс­ периментов с точечными источниками гамма-лучей. В качестве критерия формы спектра было взято эффективное разрешение фотопика и отношение впадина/пик фотопика. В качестве переменных рассматривались величины расстояния от детектора до источника, толщины рас­ сеивателя, глубины источника в рассеивателе и энергии гамма-лучей. В качестве примера достигнутых результатов приводится отношение впадина/пик для 12,5-см кристалла N aJ,

вариируюшее на фактоо ло 6 пои изменении глубины источника гамма-лучей с энергией 0,28Мэв от 0 до 2 2 см в рассеивателе с постоянной геометрией, в то время как для гамма- лучей с энергией 1,1 Мэв соответствующий фактор равняется 2. Наблюдаются связанные изменения величины эффективного разреш-ения, хотя фактор при этом не так велик. Будут представлены диаграммы, иллюстрирующие эти изменения и позволяющие произвести срав­ нения с соответствующими данными по однородно распределенным источникам.

FORMAS DE LOS ESPECTROS DE CENTELLEO. Los autores realizaron una serie de experimentos para determinar el efecto que ejerce sobre la forma de los espectros de centelleo la autodispersión que tiene lugar en el material de la fuente. Los resultados pueden aplicarse al análisis de los espectros obtenidos de personas

79 80 K. G. McNEILL and V. K. MOHINDRA

internamente contaminadas con unâ m ezcla de isótopos no identificados. Previamente se publicaron resultados obtenidos para el caso de radiactividades uniformemente distribuidas en maniquíes [1]. En la presente memoria se describen en cambio experimentos efectuados con fuentos puntiformes de rayos gamma. Los autores adoptaron como criterios de la forma espectral la resolución efectiva del máximo fotoeléctrico y la razón entre el mínimo y el máximo de la onda en correspondencia con el máximo fotoeléctrico. Como variables, consideraron la distancia detector-fuente, el espesor del dispersor, la profundidad de la fuente en el dispersor y la energía de los rayos gamma. Como ejemplos de los resultados obtenidos dan los siguientes: para una fuente gamma de 0,28 MeV, la razón mfnimo/máximo, en el caso de un cristal de Nal, de 5 pulg, varía en la proporción de 1 : 6 cuando la profundidad de la fuente pasa de 0 a 2 2 cm en un dispersor de geometría fija, mientras que para una fuente gamma de 1,1 MeV, el factor correspondiente es sólo 2. También se modifica la resolu­ ción efectiva, pero los factores de variación no son tan grandes. En la memoria estas variaciones se representan gráficamente y se las compara con las alteraciones respectivas en el caso de las fuentes uniformemente distri­ buidas.

INTRODUCTION ,

In most research with radioactive' sources, attempts are made to maxi­ mize the effects of the source itself and to minimize secondary effects due to self-absorption or scattering. In the applied field, however, one is forced often to work with extended sources or with sources embedded in a scattering material, and in both these cases the presence of absorption and scattering is unavoidable. The situation of greatest interest to the present company is that which is obtained when a source is present in the human body. In a series of experiments it has been our object to find some of the effects, source thickness does have. . From the beginning, it is clear that Compton scattering of gamma rays in the source will degrade the photon spectrum, with resultant effects in the final shape of the spectrum produced by a scintillator counter system. Figure 1 shows the effect of source thick­ ness on the shape of a mono-energetic gamma ray spectrum. Small angle scatters'in the source, or the m aterial surrounding it, w ill reduce the photon energy only a little, but will consequently distort the low energy side of photopeak, with the effect that the resolution of the system is apparently worsened: larger angle scattering may result in photons of such an energy that they will apparently "fill-in " the valley between the photopeak and the Compton edge of the scintillation spectrum: finally large angle scatters will so degrade the photons' energy that they can appear only in the Compton plateau part of the spectrum, with of course a corresponding decrease in the number of counts in the photopeak - the photofraction will have been effec­ tively decreased. ' • . Now in many cases, such as the long-term following of the decay of . an isotope for physiological research, such effects may not be of importance. It is, however, possible to think of situations when changes in photofraction and resolution due to the source itself may be of importance - for instance, in the identification of unknown isotopes in a person as a. result of accident, it is important to know if an unusually wide photopeak is due to the presence of two isotopes' emitting gamma rays with nearly equal energies, or simply due to the obeseness of the subject. Again, an apparently low photofraction may be due to the presence of a low energy gamma ray or to a large amount of self-scattering in the source. We have therefore made studies of the effect of source thickness on .scintillation spectra shape. SHAPES OF SCINTILLATION SPECTRA 81

Fig. 1

Diagram to illustrate the general effects that occur when a thick gamma source is viewed by a scintillation counter. '

EXPERIMENTS

In all our work to study these effects, we have used water phantoms of variable depths to simulate scattering in the body. In the first series, by myself and R. M. Green, active material was dissolved uniformly in the water, and variations in shape studied as a function of depth of the source. This work has already been published [1]. A 5-in diam. X4-in thick- Nal(Tl) crystal was used as the detector, and the'phantom was a box of area 41 cmX34 cm, which could hold liquid up to a depth of 44 cm-. Six sources, with energies ranging from 0.32 MeV to 1.53 MeV were used in the experiments. As criteria of the changes in shape produced, we used in particular the relative changes in the apparent resolution Of the photopeak and the relative changes in the valley-to-peak ratio as the thick­ ness of the source, that is the-depth of active solution in the-phantom, was increased from zero. The results obtained are illustrated in the next two figures. In the first (Fig. 2) is shown the effect of thickness on apparent resolution. The maxi­ mum effect is obtained with the lowest energy gamma ray used, 0.32 MeV. Here, in absolute term s, the resolution altered from 16.3% with a point source of Cr5i to 20.0% when the Cr was uniform ly dissolved in water of depth 33 cm, that is a change of 23%, as shown on the graph. Such a change is consistent with the results of ;simple calculations considering only small angle scat­ tering within the so.urce. For the high energy gamma rays the effects are smaller, amounting to a change of only 5% in the ¿ase of the 1.53 MeV gamma rays from K42. . ■ Figure 3 shows the changes in the valley-to-peak ratios for the same sources and phantom depths, the valley-to-peak ratio being defined as the 82 K. G. McNEILL and V. K. MOHINDRA

F ig . 2 -

The relative change in the apparent resolution of a photopeak when the thickness of a uniform source is increased. lowest count, in any channel between the photopeak and the Compton edge divided by the highest count in the photopeak. Here the relative changes are much larger, instead of a 23% change, the 1 valley-to-peak ratio for Cr51 changes by over a factor of 3 as the source depth increases from zero to 33 cm, and even with 1.53 MeV gamma rays the change is the order of 25%. Correlation curves may be drawn enabling one to.predict the expected resolution from the measured valley-to-peak ratio1'. The above situations are those analogous to the whole-body measure­ ment of someone with, for instance, caesium-137 uniformly distributed throughout the body. In a second series of experiments we have investi­ gated what happens to the shape of scintillation spectra when a point source is immersed in a scattering medium, which corresponds to the presence of a localized source in the human body. -­ Once again a 41 cmX34 cmX44 cm tank was used, water again being the scattering medium. As detector", another 5 inX4 in Nal(TI) crystal was employed. The experimental arrangement is shown in Fig. 4. Even with the detector at a fixed distance from the bottom of the phantom (H constant), there is the possibility of an infinite number of different combinations of the distance D of the source below the surface of the phantom and the total depth h of the water sample. This is in contradistinction to the situation with the uniformly distributed sources, where h was the only variable con­ sidered. For simplicity we shall here limit ourselves.to the case where the thickness of the scattering material remains constant and only the distance of-the source below the surface of the water is varied. This corresponds to differing locations of a point source in one individual. ' . Once again the shape of the scintillation spectrum is changed in the same general way. As the source moves further and further below the sur- SHAPES OF SCINTILLATION SPECTRA 83

Fig- 3

The relative increase of the valley-to-peak ratio (the lowest count in any channel between the photopeak and the Compton edge divided by the highest count in the photopeak) with increasing thickness of a uniform source.

A ♦ photomultiplier

Nal (T O

Fig. 4

The experimental arrangement for studying the effect on spectrum shape of varying the depth of a point source in a water phantom. face of the water scatterer the effective resolution worsens and the valley between photopeak and ¡Compton edge becomes filled up, with a correspon­ ding increase in the valley-to-peak ratio. These effects, as in the case of 84 K. G. McNEILL and V. K. MOHINDRA the uniform distribution of activity in the scatterer, are due to small angle scattering. ■ ' The point source consisted of, a small'spherical glass bulb (1 to 1.5 cm in diameter) containing about 5 дс of-radioactive solution. The source was supported on a lucite platform. For D less than 12 cm a similar source, but of 1 Цс, was used in order to.reduce count rate dependent shifts of the photomultiplier gain. It should be noted,’ then, that strictly the source was not a "point".

. , Fi8 - 5 .

The relative change in the photopeak resolution .as’sources of different energy are submerged to different depths in a’-water phantom.

Figure 5 shows the relative worsening of the photopeak resolution with increasing submersion of the point1 source. • Comparison with the corres­ ponding results for uniformly distributed sources (Fig. 2) at once shows some marked differences (although caution must be used in making detailed inter­ comparison because the curves obtained are functions of the crystals them­ selves). For instances with Cr 51 gamma rays (0.32 M eV) from a point source 2 0 cm below the surface of the water, the effective resolution is more than 50% worse than that obtained with a free point source, whereas with Cr5-i uniform ly distributed in water of depth 2 0 cm, the resolution only worsened by about 20%. - The shapes of the two sets of-curves are also different with the submerged point source, the resolution apparently continues to worsen with increasing depth, while with the uniformly distributed sources (Fig. 2) SHAPES OF SCINTILLATION SPECTRA 85 the rate .of change of effective . resolution seems to flatten off at larger depths. . . ' . These differences must be explainable in terms of the fact that the uni­ formly distributed source consists essentially of an infinite number of point sources and so, to a first approximation, it might be expected that the reso­ lution for the case of the uniformly distributed source of depth 2 0 cm would correspond to that for the point source at 10 cm. This in fact is not far from the truth. To carry the explanation further, one must realize that both true absorption of gamma rays in the water medium and geometrical considerations w ill weight the-effect of activity in-the upper layers qf the liquid, and thus to this approximation the apparent resolution of the photo­ peak due to gamma rays from a. uniformly distributed source of depth x cm should be equal to that obtained with a point source situated at rather less than x/2 cm below the surface of the liquid. This is what is found when x is sufficiently large that the apparent resolution curve has begun to flatten out.

Fig. 6

The relative change in valley-to-peak ratio for point sources at different depths in a water phantom.

Similar results are found for the valley-to-peak ratios (Fig. 6 ). O ver the source depth studied, there is a continuing, indeed nearly a linear, rise of the ratio with source depth, the rate of rise being greatest for the lowest energy gamma rays. Once again the changes are much greater than those found with uniformly distributed sources. It is our hope that the fact that such relationships between resolution, valley-to-peak ratio and source thickness exist will be of help to users of whole-body counters in the analysis of scintillation spectra obtained frpm 8 6 K. G. McNEILL and V. K. MOHINDRA persons with a radioactive burden - we would endorse the suggestion made by RUNDO [2] that the fact that such large changes occur in the valley-to- peak ratio with depth of source below the surface of the scatterer might be used as a quick way of determining the location of a point source using only a single Nal(Tl) crystal.

■ REFERENCES '

[1 ] McNEILL, K. G. and GREEN, R. M. , C añ ad. J. Phys. 39 (1 9 6 1 ) 1 8 4 2 -5 0 . ■ [2] RUNDO, J ., Proc. 2nd. UN Conf. PUAE 23 (1958) 101. 87

DISCUSSION : (on the three foregoin g papers)

Y. NAVERSTEN: I was very interested in the results discussed by Professor McNeill in the last paper. I obtained similar results working with three nuclides in different parts of a plastic Alderson phantom. I also found considerable differences in the shapes of the spectra for I131, Cs132and Ca47 in different parts of the phantom, especially in the low-energy part. With regard to the sentence commencing "With the crystal 30 cm above the surface ..." in the seventh paragraph of the paper "A new technique for determining the distribution of radium and thorium in living persons", I should like to ask Dr. M iller whether this also applied to low-energy photons in the 75 - 275 and 275- 375 keV energy bands. And I wonder if it would not be better to use somewhat narrower energy bands at the low-energy end of the spectrum. ■ C.E. MILLER: The counting rates in all five energy bands remain es­ sentially constant as the phantom thickness is increased from 8 cm to 2 0 cm. The counting rate observed in the 75 - 275 keV energy band does not change as would be expected from the data presented in the paper by Professor McNeill. This is probably due to the fact that radium emits X-rays and gamma rays of 73 keV, 188 keV and 241 keV which fall within this energy band. I should like to take this opportunity of asking Mr. Naversten a question about the paper "A two-crystal scanning bed counter for accurate determi­ nation of whole-body burdens". Is the crystal moved at a uniform speed along the length of the patient and the resultant spectra simply accumulated in the memory of the analyser, or are spectra obtained at either one or several different positions along the body? Also, since the counting rate observèd along the length of the body from a uniformly distributed quantity in the body w ill increase as the crystal is moved from the head toward the legs, the sensitivity of the crystal ef­ fectively changes as the crystal is moved along the body. How do you correct for this changing sensitivity? Y. NAVERSTEN: Norm ally, when we use this technique for the determ i­ nation of whole-body activity, we use only the sum of pulses from a full scan without correcting for. changing sensitivity or for different points in the geometry. The geometry itself smoothes out the effect of varying distri­ butions. At the subject's head scanning efficiency using a С si37 point source is only 50% of what it is at the middle of the scan. Correction for different efficiencies can be obtained by altering the orbit of the detectors or by using an arc instead of the orthodox stretcher and measuring the subject in both a prone and a supine position. We are at present developing this method. W.V. MAYNEORD (Chairman): In my own experiments I found that, if you took a monochromátic source, the mean energy dropped as you moved away from or into a large scattering ma.ss. With a sufficiently heterogeneous source the filtration effects might first cause softer and then harder radi­ ation, or possibly a. complete, immediate fall-away. I think that with the right geometry and combination of isotopes you could have all three effects, two or three of them at the same time. There is probably no simple rule in this matter; it will depend on the geometry and the conditions. 88 DISCUSSION

С. J. MALETSKOS: I should like to ask Mr. Naversten a series of questions. My first question relates to Fig. 2 in his paper, from which it appears that the length of the subject was nearly the same as that of the room . Would not a scan of "in fin ite" length be better than a lim ited-length scan, i. e. would it not be better to begin scanning at a point w ell before the head and continue beyond the feet? Y. NAVERSTEN: If you make the'scans longer the counting efficiency will decline. As I showed in Fig. 8 , the counting efficiency is approximately in inverse proportion to the scanning length.’ Also, if you make the scanning length longer, the oblique passage of photons through the subject will result in a more pronounced variation of the counting efficiency in the posterior- anterior direction. An infinite scan would reduce only the geometrical vari­ ations. Variations due to absorption effects, which are dominant, would not be-reduced. . С . J. MALETSKOS: My second question also relates to the size of the room. How do changes in the magnetic fields within the iron room affect the scan? • Y. NAVERSTEN: We have not observed any effects arising from changes in the magnetic fields within the iron room. C. J. MALETSKOS: In our iron room we have variations of at least 10 to 15% depending on how close the crystal comes to the wall. And if it starts from a point very close to the wall and moves out we get changes not only in spectrum shape but also in counting rate. Y. NAVERSTEN: Counting in our iron room has never been performed with the vertical axis of the crystal closer than 35 cm to the wall. The geometry was developed in an iron room but will now be used in a newly constructed lead room. C. J. MALETSKOS: Thank you. Lastly, with scanning carried out con­ tinuously, ■ that is with a power drive, how do you make live-tim e corrections as the detector passes over non-uniform distribution within the body? Y. NAVERSTEN: Firstly, we always use very small counting rates. Secondly, we find no difference in results, whether we correct for dead time in different parts of the scan or divide the number of counts obtained during the scan by the measuring time and correct by using that figure. J. RUNDO: Does Mr. Naversten have any evidence of variations in counting efficiency with variations in subject height? Has he considered the possibility of making the scan a constant fraction of the subject height? With a scanning length of 144 cm there must surely be considerable differences between results for persons 140 cm and 200 cm tall. ■ Y. NAVERSTEN: A fairly large group of patients were injected intra­ venously with solutions of S rS 5 (about 0.5 ц с). The patients were measured within 30 to 60 min, when most of the activity was in the circulatory system. The results show that the response might vary by about 2% per 10 kg of body weight. If this is true there is probably also a variation with length of the subject. ■ The arrangement in the iron room allowed a scan of only 144 cm. In the lead room (in operation since last year) the scan can be as long as 185 cm. For routine work the technique has to be rather standardized? Until now no investigations have been perform ed on a scanning length correlated to the length of the subject. ■ . DISCUSSION 89

F ig . I D .

Whole-body counter erected at the National Institute for Oncology at Budapest

L. BOZÓKY: I should like to give some details of our whole-body counter, erected in the National Institute for Oncology in Budapest and at present being used for the examination of Ra- and'Th-contaminated persons ^Fig. ID). The apparatus consists of an iron shield, the walls of which are 2 0 cm thick and covered with Pb- and-Cu-sheets, and of a plastic scintil­ lator 26.5 cm long and 30 cm in diameter. The interior dimensions are 60 cm X 170 cm X 100 cm (height). The subject sits in a "standard-chair" which rolls out automatically when the door is opened. Filtered air is passed continuously through the chamber. A built-in loudspeaker provides entertainment for the subject so as to prevent claustrophobia. - ' The scintillator is coupled by means of a Plexiglas light pipe to a 12-cm EMI photomultiplier, and the pulses are recorded by a single-channel ana­ lyser.^ Five counts per second correspond to a phantom containing 150 g of potassium in the "potassium region" of the spectrum.

HUMAN BETA BREMSSTRAHLUNG DETECTION BY MEANS OF THIN AND THICK SODIUM IODIDE CRYSTALS

L. G. BENGTSSON RADIATION PHYSICS DEPARTMENT, UNIVERSITY OF LUND, SWEDEN

Abstract — Résumé — Аннотация — Resumen

HUMAN BETA BREMSSTRAHLUNG DETECTION BY MEANS OF THIN AND THICK SODIUM IODIDE CRYSTALS. Thin Nal(Tl) crystals have earlier been used in attempts to assess the body burden of beta-emitting nuclides in the human body by means of the bremsstrahlung produced in the body. In many respects large Nal(Tl) crystals are more advantageous. The two detector types are compared regarding their statistical properties and their performance when large amounts of interfering gamma emitters are present in the body of the measured subject. The use of 42-cm chair geometry versus 35-cm and 45-cm scanning bed geometry is discussed. An account is given of the influence of body weight and shape on the background spectrum and on the spectra of Cs137 and K40. Possible errors in the assay are discussed.

Results obtained from measurements of several subjects with an 8 in x 4 in crystal in the scanning bed geometry are reported. The excess count in the energy interval used for bremsstrahlung assay was in all cases, if no unknown contamination was present, less than the count corresponding to 3 SD (theoretically obtained standard deviation). The minimum detectable amount of any beta emitter, using the reported method, is thus expected to be that which corresponds to 3 SD. A 2-h measurement shared on subject and background would then reveal a contamination of about 40 nc Sr90 + Y^° in a subject containing 50 nc Cs*37. .

. DETECTION DU RAYONNEMENT DE FREINAGE DES ÉMETTEURS BETA DU CORPS HUMAIN, A L’ AIDE DE CRISTAUX MINCES ET ÉPAIS D'IODURE DE SODIUM. Pour déterminer, à l'aide du rayonnement de freinage produit dans le corps, la charge corporelle de nucléides émetteurs bêta, les auteurs avaient antérieure­ ment utilisé des cristaux minces de N al(Tl). A bien des égards cependant, de grands cristaux de Nal(Tl) sont plus avantageux. Les auteurs comparent les deux types de détecteurs du point de vue de leurs propriétés statistiques et leur rendement dans les cas oti de fortes quantités d’émetteurs de rayons gamma interférents sont présents dans le corps du sujet. Ils comparent aussi les avantages respectifs de la géométrie de la position assise (42 cm) et ceux de la gammagraphie en position allongée (35 et 45 cm ). Ils décrivent les effets du poids et de la forme du corps sur le spectre du bruit de fond, ainsi que.sur les spectres de 137Cs e t40K et examinent les erreurs qui peuvent être introduitës dans le dosage. Les auteurs présentent les résultats relatifs à plusieurs sujets qui ont été mesurés à l’aide d'un cristal de 20 x 10 cm dans la géométrie de la gammagraphie en position allongée. Dans tous les cas - en l'absence de toute contamination inconnue - le nombre de coups en excédent, à l'intérieur de l'intervalle d'énergie utilisé dans le dosage à l'aide du rayonnement de freinage, était inférieur au nombre correspondant à 3 ET (écart type théorique). En recourant à-la méthode indiquée,' il faut donc s'attendre que la quantité minimum décelable de n'importe quel émetteur bêta sera celle qui correspond à 3 ET. Une mesure de deux heures partagée entre le sujet et le bruit de fond révélerait alors une contamination d’environ 40 nc de 90Sr et 9ÛY chez un sujet ayant une charge de i37Cs égale à 50 nc. . '

ОБНАРУЖЕНИЕ ТОРМОЗНОГО БЕТА-ИЗЛУЧЕНИЯ У ЧЕЛОВЕКА С ПОМОЩЬЮ ТОН­ КИХ И ТОЛСТЫХ КРИСТАЛЛОВ ЙОДИСТОГО НАТРИЯ. Ранее в целях-определения бета- излучающих изотопов в организме человека, благодаря тормозному излучению, образующе­ муся в организме человека, использовались тонкие кристаллы N aJ(Tl). Во многих отноше­ ниях крупные кристаллы N aJ(Tl) более удобны. Сравниваются два вида детекторов по их статистическим свойствам и их работе там, где в организме исследуемого объекта имеются большие количества интерферирующих гамма-излучателей. Обсуждается вопрос об исполь­ зовании геометрии 42-см кресла наряду с 35 и 45-см скеннирующей геометрией койки. Уде­ ляется внимание влиянию веса организма и формы на фоновый спектр и на спектры цезия-137 и калия-40. Обсуждаются возможные ошибки при определении.

91 92 L G, BENGTSSON

Приводятся результаты, полученные от измерений нескольких объектов с помощью крис­ талла 20 X 10 см в скеннирующей геометрии койки. Во всех случаях превышение счета ин­ тервала энергии во время определения тормозного излучения, если отсутствует неизвестное загрязнение, было меньшее, чем это может соответствовать 3 SD (теоретически полученное стандартное отклонение). Таким образом, используя вышеизложенный метод можно ожидать, что минимально обнаруживаемые количества любого бета-излучателя окажутся соответствую­ щими 3SD . Двухчасовое измерение объекта и фона показало радиоактивное загрязнение почти в 40 ммк кюри стронция-90 и иттрия-90 в объекте содержащем 50 ммккюри иезия-137.

. EMPLEO DE CRISTALES DE YODURO DE SODIO DELGADOS Y GRUESOS PARA LA DETECCIÓN DE EMISORES BETA EN EL CUERPO HUMANO POR RADIACIÓN DE FRENADO. En los anteriores intentos de evaluar la carga-corporal de emisores beta en el hombre por la radiación de frenado emitida en el cuerpo se empleaban cristales delgados de N al(Tl). Ahora bien, los cristales grandes de Nal(Tl) son más convenientes desde muchos puntos de vista. Los autores comparan las propiedades estadísticas de ambos tipos de detectores y su rendimiento cuando en el cuerpo existen cantidades grandes de emisores gamma que pueden interferir. La memoria compara el empleo de una geometría de silla de 42 cm con el de geometrías de camilla de exploración de 35 y 45 cm, e informa sobre la influencia del peso y la forma del cuerpo en el espectro de fondo y en los espectros del i37Cs y d e l 40K. También se discuten los errores que pueden producirse en el análisis. Se informa también sobre los resultados obtenidos al medir a diversos sujetos con un cristal de 20 x 10 cm en una geometría de cam illa de exploración. En todos los casos en que no había presente ninguna contami­ nación desconocida, el exceso del recuento en el intervalo de energía empleado para analizar la radiación de frenado era inferior a la cifra correspondiente a 3 DS (desviación standard obtenida teóricam ente). Por tanto, si se emplea el método descrito cabe esperar que la cantidad mínima detectable de un emisor beta sea la que corresponde a 3 DS. Así, pues, una medición de 2 h de duración repartida entre el sujeto y el fondo revelaría una contaminación de unos 40 ne de 90Sr y 90Y en un sujeto que contuviera 50 nc de 137C s.

1. INTRO D U CTIO N

Measurement of external bremsstrahlung from pure beta emitters in man has many applications [1, 2, 3, 4, 5, 6 ] . Detection of Sr 90 has a wide interest in the control of excisable body burdens due to fall-out. In medical therapy and diagnosis pure beta emitters such as P 32 and У 9° are infrequent use. However, whole-body counting as a tool for their exploitation has so far been poorly investigated. , In order to make full use of the possibilities offered by bremsstrahlung, reliable methods for the assay of beta emitters in the whole body, or parts of it, have to be worked out. The effect of body weight, localization of the activity in the body, etc. on the detector sensitivity must be investigated. Special studies of the minimum detectable amount of pure beta emitters in the huynan body have to be performed with particular reference to determi­ nation of body burdens of (Sr+Y ) 90 from fall-out and reactor accidents. Im­ provements in the detection of small amounts of beta emitters also give the possibility of decreasing the amounts of radionuclides necessary for medical diagnostic studies. To achieve good detectability the proper "background" for the bremsstrahlung spectrum studied must be carefully investigated.. This background consists of contributions from the radiation field around the object and detector, and from radionuclides present in the body of the subject, such as Cs137 and K 40.

2. REVIEW OF EARLIER WORK

In a paper by LIDÉN and McCALL [8 ] a special low-energy-photon de­ tector, suitable for whole-body counting, is described. They used an as­ BETA BREMSSTRAHLUNG DETECTION 93

sembly of two Nal(Tl) crystals, 5-mm thick and 110 mm in diameter, with a 0. 25-mm Al-window. With these crystals the bremsstrahlung from Y 90 in a human subject was investigated. An attempt was made to use thin crystals for the detection of very small amounts of beta emitters in the human body. The technique used consists of a comparison of a subject's measured spectrum in the low-energy region with a synthesized spectrum, obtained from the general background around the detector and its change due to the presence of the object near the detector, and from scattered radi­ ation as w ell as Compton distribution in the crystal accompanying the gamma photons emitted by radionuclides in the body of the subject, usually Cs 137 and K40. The difference between the measured and the synthesized spectrum is then due to radiation from other radionuclides in the subject. If no corre­ sponding photopeaks are found, these nuclides are very likely to be pure beta emitters and the remaining spectrum is due to their .bremsstrahlung. The scanning bed geometry was preferred for the thin crystal measure­ ments. The background was measured with sugar bags simulating a subject. Through phantom measurements, the contribution in the low-energy band per unit activity of Csi37 and K40 was determined. The subject's content of these nuclides was found by a measurement with a large crystal in a chair geom etry.

3,- EFFECTS OF LARGE AMOUNTS OF INTERFERING GAMMA EMITTERS

The ordinary background in the iron room at Lund in the energy band used, 28-160 keV, was about 140 cpm (counts per minute) with the thin crystals. 100 g К in the subject contributes 24 cpm; 100 ne Cs137 contri-

TABLE I . .

INFLUENCE OF AN ERROR IN THE DETERMINATION OF THE C s 137 CONTENT (SD = 3%) ON THE MINIMUM DETECTABLE AMOUNTS OF Sr 90 IN THE SUBJECT (MDA)

The subject is measured for t min in the scanning bed geometry with the thin crystals; sugar background is also run for t min. A potassium content of 120 g is assumed.

. MDA (n c Srs°)

c s ls7 : negligible error in 3% SD in .

content C s 137 determination C s -37 determination

(n c) t = 2 0 m in t = 40 m in t = 2 0 m in t = 4 0 m in

0 1 34 24 34 2 4

■ . 5 35 25 3 5 ■ 25

1 0 35 25 36 26

50 ' 41 29 54 4 5

1 0 0 4 7 . 33 83 77

500 79 56 356 3 5 4

1 0 0 0 106 . 7 5 ' 701 700 94 L. G, BENGTSSON butes 260 cpm. It is evident that for subjects containing hundreds of nano- ' curies Cs137 this radionuclide will dominate the subject's gross spectrum in the bremsstrahlung energy band. It is also obvious that in this case the accuracy of the spectrum synthesis discussed in section 2 is the same as the accuracy with which the Cs 137 contribution can be determined. Measure­ ments with the thin crystals on several subjects containing 166-815 nc Cs137 revealed a SD (standard deviation) of about 3% for the measured sensitivity of the counter for Cs 13.7 contribution in the bremsstrahlung band (cpm in 28-160 keV per nanocurie Cs137). This is mainly due to the inaccuracy in the chair geometry determination of the Csi37 content. Table I shows how this influences the obtainable minimum-detectable-amounts, defined as the amount of Sr9o corresponding to two SD in the measurement. For more than 50 nc Cs137 the effect is considerable. With increasing Cs137 content prolonged measurement times offer no advantages, and for very high C s 137 contents the minimum detectable activity of Sr90 tends to ap­ proach a factor 0.7 of the activity of Cs137, regardless of measurement time. At 1000 nc the deterioration is a factor of seven for a (20+20) min measurement. ‘ .

4. USE OF A LARGE CRYSTAL ' .

According to the technique mentioned in section 2 the subject is measured twice in different geometries. Inevitably, errors are introduced due to the limited reproducibility of the geometries involved. In the case of subjects with high gamma activity an important reduction of the errors would result if only one measurement were employed. This implies the use of a crystal that combines a high photopeak sensitivity for the inter­ fering gamma rays with a low "background" in the bremsstrahlung energy band. Since high photofraction also means that the Compton tail is small, the advantage of a large crystal, giving high photofraction, increases when the "background" is mainly due to radionuclides in the body of the subject. A large crystal has been used by GOLDMAN, YOUNG and EDMONDS [9] for whole-body counting of Sr90 in beagles. The use of a large crystal puts certain requirements on the equipment. The general equipment will now be described in detail.

5. E Q U IP M E N T

For the determination of small radioactive burdens in man different whole-body counting geometries have been developed. The high sensitivity has made the chair geometry very common. However, this geometry is sensitive to changes in the activity distribution in the body and to varying body size and shape, more so than the scanning bed geometry used in the whole-body counter at Lund since 1961. .Therefore, in this investigation interest has been focused on the scanning bed geometry, although some com­ parison with the chair geometry is made. In the bed geometry the patient is measured both in the prone and supine positions on a stretcher, while the detector scans above the bed at a distance of 35 cm or 45 cm from the centre of the plane surface of the detector to the bed. BETA BREMSSTRAHLUNG DETECTION 95

The thin crystals were described in detail in [ 8 ] . Since that time the photomultipliers have been replaced by EMI tubes, which give a background reduction of 40%,in 28-160 keV band. The thick crystal used is an 8 X 4 in Nal(Tl) crystal enclosed in 0.5 mm stainless steel. The MgO reflector thickness on the flat surface is 0.5 g j cm ¿ . Three 3-in Dumont photomulti­ pliers are used and the optimum resolution for the Cs13" 662-keV gamma line is 9.8%. To obtain best stability the high voltage is kept at 1000 V. Via a current integrator type of pre-am plifier the pulses are fed to a RIDL 200 channel analyser. With the 8 X 4 in crystal this anályser w ill be used for a spectrum extending from the low-energy photon cut-off (due to the iron housing) at about 40 keV to the upper end of the K40 peak at about 1600 keV. Most analysers have a discriminator, which limits the low- energy end of the spectrum, and an amplifier saturation limiting the upper, end. For the RIDL analyser, the discriminator level after the amplifier is given as 0.16 V and the amplifier saturation begins at 8 V. Theoretically, with 200 keV per volt it would be possible to cover the range 32-1600 keV. In practice the lower limit must be set at 56 keV and still a slight non­ linearity above 1300 keV has to be accepted.

6 . SPECTRAL DISTRIBUTION OF BREMSSTRAHLUNG

6.1. Sources of bremsstrahlung

The nuclide most used as a bremsstrahlung source is Y90, which decays with a half-time of 62 h to stable Z r 90 with emission of beta rays having a maximum energy of 2. 24 MeV. The well-known beta emitter Sr 90 with h alf-tim e 28 y r and E max = 0. 54 M eV decays to Y 9o. in this investigation the Sr90 used as a bremsstrahlung source was always in equilibrium with its daughter product Y 90. In this section a point bremsstrahlung source is mentioned. In this source bremsstrahlung is produced by 10 цс Sr90( + Y90) in plaster of Paris (CaS0 4 - 2 H 2 0 ) thick enough to absorb almost all of the beta rays. The plaster of Paris diameter was 36 mm and the entire thickness of the source and its mounting was 23 mm. . In one case bremsstrahlung from P 32 in a human subject was measured. The maximum beta energy of the P 32 beta decay is 1. 7 MeV.

6.2. Measurement of bremsstrahlung spectra

The point bremsstrahlung source was measured with both the thin crystals and with the thick one, with the detector in a fixed position and a 42-cm distance from the crystal surface to the source. The result is shown in Fig.l. The thin crystals detect bremsstrahlung down to 30 keV, while the detection with the 8 X 4 in crystal is impaired below 60 keV. The large crystal is of course more efficient at higher energies. (Normalized to equal crystal surface areas the counting rates are the same at about 120 keV. At this energy the iron housing and the magnesium reflector of the 8 X 4 in crystal absorb 17% of incident photons while 16% do not interact in the 5-mm thick crystals. ) • 96 L. G, BENGTSSON

, '

Spectra of a Sr™. point bremssirahluni; source. ' ++ Thin crystal assembly; 2 crystals, 11-cm diam. ; front surface area, 190 cm'.

oo 8X4 in crystal; 8 -in diam; front surface area, 324 cm2. Bars indicate ± 1 SD. .

The bremsstrahlung source was also put under 4 cm and 7 cm of water in a container with a cross-section area of about 600 cm2. The spectra were m easured but no great change in the spectrum shape was found, as appears from Fig. 2. Note that at lower energies the counting rate is even higher with 4 cm of water than without.

F ig . 2

Spectra of a Sr90 point bremsstrahlung source put under 0, 4 and 7 cm of water (8x4 in crystal).

In Fig. 3 the spectrum of the source under 7 cm of water is compared with the spectrum obtained from a human shaped phantom in scanning bed geometry, having Sr90+Y9<> dissolved in a 22% solution of СаС1г in its femur, and with a spectrum from a human subject three days after injection of Y 90 [8 ]. For all spectra the thin crystals were used. The most apparent anomaly is that for the spectrum from Sr90+Y9° in bone-equivalent solution BETA BREMSSTRAHLUNG DETECTION 97

Fig. 3

Spectra of a Sr 90 point bremsstrahlung source under 7 cm of water,

and from scanning bed measurements of Y 90 in a human subject and Sr90 in the femur of the Alderson phantom. Thin crystal assembly. Normalization to maximum value is made. The spectra of the point source and the human subject overlap below 60 keV.

F i g .4

Bremsstrahlung spectra obtained with the 8X 4 in crystal.

The spectra were obtained from a measurement of the St 90 point bremsstrahlung source

under 7 cm of water and from scanning bed measurements of Sr90 solution

in the femur of the Alderson phantom and P32 in an obese patient. Normalization to the maximum points of the spectra is made. in the phantom femur; in that case the low-energy cut-off starts at 10-keV higher energies than for the two other spectra. A comparison of spectra 98 L. G. BENGTSSON from a point source under 7 cm of water and a phantom femur measured with an 8 X 4 in crystal reveals a similar difference of about 8 keV. This is shown inFig.4, where a spectrum from P 32 in an obese patient is also shown. The low-energy cut-off in this spectrum also starts at higher energies than in the spectrum from a point source under water. No such difference was found by LIDÉN and McCALL [8 ] when Y 90in the femur was measured with the thin crystals fixed 35 cm over the bed (measuring the single thigh) and compared with the spectrum from the measurement of Y 90 in a human. Another difference between the shapes of the three spectra in Fig. 3 is that the spectrum corresponding to the measurement of Y 90 in a human contains relatively more photons with energy above about 150 keV. This might partly be explained by the presence of bremsstrahlung radiation from the low-energy beta particles due to Sr90 decay (Emax =0. 54 MeV) in the two other spectra.

7. DETERMINATION OF OPTIMUM ENERGY INTERVAL FOR ASSAY OF SMALL AMOUNTS OF BETA EMITTERS

The spectral shape is related to the determination of a suitable energy interval fo r the assay o f sm all amounts of beta em itters in the human body. "Sm all" means that, in this interval, the net counting rate, S, from the beta emitter is small compared with the "background", 3 , mentioned in section 1 , equivalent to the synthesized spectrum defined in section 2 . We want a figure of merit to describe the relative statistical goodness of the system when some parameter (e.g. the energy.interval) is changed. Let В be the counting rate contribution from the radiation field around object and detector in some bremsstrahlung assay interval ДЕ. Let the gamma emitting nuclides A2, A3., .... give counting rates R^ R2, R 3 ...... in suitable energy intervals covering their respective photopeaks. Corre­ sponding energies are arranged with Ei< E 2 < E 3 < .... Fractions fi, f2, f3, .... of these counting rates contribute to the "background" j3 in the inter­ val for the bremsstrahlung assay. "Background" counting rates in the photo­ peak intervals are denoted j3j, /32 , /З3 , .... The measuring time of the subject is ts and of the background t'e . We assume that the e rro rs in the value of the fractions fj, f2, f3, .... are negligible. The standard deviation crn in /3n is complicated. We do not need the full expression for our considerations but assume that to it can be attributed a given value. The variance in the synthesis of 0n is thus a n2. The variance in the measured gross counting rate, in peak number n, R n+|3n, is (R n+|3n)/ts . The variance in Rn is thus (Rn+i3n)/ts +<7$. A n contributes f nR n to the

counting rate in ДЕ and the variance in this contribution is f ? (—Ir— —0 + crj;). . . ts ' Now, we get a rest spectrum in ДЕ when we subtract all contributions E fnRn

and В to (3 from the measured counting rate |3 + S. We assume that S

' • V= —. + — + £ Íü ib :+z;f 2 (5п1Ап+ст2'\ ' (1) . ts tB П ts n П V t s n' '

where we have used /3 = В + £ fnR n . • Two special cases have great interest: • (a) Proof exists that no gamma contamination is present. In this case we obtain

V-“ . “ . ■

Note that even if Rn = 0, there must be good reason to assume an uncontami­ nated subject, otherwise the variance is increased by the last term of Eq. (1) with Rn= 0. . With a given time T available, so that ts +te = T, differentiation of Eq.(2) with respect to time shows that the smallest variance is achieved for ts = tj = T/2.. Thus we have V = 4B/T. The relative variance is 4B/S2T and the inverse of this is used as a figure of m erit for evaluation of the opti^ mum values of various parameters. The constants then have no significance so in this case we use as a figure of m erit S2/B.

(b) One gamma emitter dominates the spectrum. Let.this gamma emitter be A j. The following approximations are valid:

Rj » Rk for к / 1 f i R l » fi(Rkfork/l • f j R - t » В ■ R i » Pk fo r all к

For simplicity we may omit the index 1. With these approximations we have

' v = - 5 - - 3 Æ *в ts • ts

As for Eq.(2) we differentiate Eq.(3) to find the best way to share a given time T on te and tg . We find the solution tg = TBi/(fR+f 2 R )i . This gives V = fR(l+f)/T. The relative variance is. fR(l+f)/S 2 T . This should be compared with 4B/S2T from case (a). With f = 1 we have won a factor two since we need hardly use any.time for the-measurement of B. As-a figure of'merit for this case we choose S2/fR(l + f). ■

(c) .Applications. The figure of merit in (a), S2/B, is valid as good approxi­ mation for the .measurement of a subject with no other gamma contamination than the inevitable potassium. . Table II shows typical calculations of Sz/B for.the thin crystals and for the 8 X 4 in crystal. It m ay be worth pointing, out that the figures are only fo r .use relatively. Thus they are valid although it happens that the units have been chosen so that S as defined in the table 100 L. G. BENGTSSON

TABLE II

STATISTICS OF BREMSSTRAHLUNG MEASUREMENTS S = net cpm from 'a measurement of the point bremsstrahlung source under 7 cm of water at a 42-cm distance. B = background cpm in the same geometry. Figure of merit is S2/B.

Energy S2/B interval S В (keV ) (X 1 0 '3 )

8 x 4 in 3 6 - 80 32 9 8 87 154 crystal 3 6 - 1 2 0 5738 175 188 3 6 -1 6 0 7178 251 2 05

3 6 - 2 0 0 80 9 0 326 2 0 1 3 6 -2 4 0 8701 401 189

2 0 -1 6 0 7610 275 2 1 0

2 8 -1 6 0 7445 262 2 1 1 3 6 -1 6 0 7178 251 205 4 4 - 1 6 0 6753 239 191 5 2 -1 6 0 62 9 6 224 171 5 6 -1 6 0 5881 215 161

Thin 2 8 - 80 3632 58 227 crystals 2 8 -1 0 0 46 1 2 •84 253 2 8 -1 2 0 5242 103 267

2 8 -1 4 0 5652 12 1 264 2 8 -1 6 0 '5912 136 258 2 8 -1 8 0 6080 150 247 2 8 -2 0 0 6185 162 236

2 0 - 1 2 0 5563 1 1 2 276 2 4 -1 2 0 ■542 9 107 276 2 8 -1 2 0 5242 103 267

3 2 - 1 2 0 5014 1 0 0 251 3 6 -1 2 0 4753 97 2 3 4 4 0 -1 2 0 44 7 4 93 216t 4 4 -1 2 0 411 2 39 181*

is much greater than B. The table shows that there is no practical use in extending the interval for the bremsstrahlung assay higher than to 200 keV for the thick crystal and to 140 keV for the thin crystals. The figure óf merit discussed in (b) is valid in the measurement of a subject having hundreds of nanocuries of Cs137 as the only contamination besides his potassium. Table III shows a comparison of the proper figure of merit, S2/fR(l+f) and S 2/fR, which is the analogue of S2/B. It is obvious that the proper figure of merit favours narrow energy intervals, as could be expected from its l/ (l + f) dependence. BETA BREMSSTRAHLUNG DETECTION 101

TABLE III •

STATISTICS FOR A BREMSSTRAHLUNG MEASUREMENT W ITH THE 8 X 4 in CRYSTAL ON A SUBJECT WITH HIGH Csi37 CONTENT For technical reasons the lower limit of the energy interval cannot be reduced below 56 keV although this would improve the figure of merit. S =net cpm from a measurement of the point bremsstrahlung source under 7 cm of water at 42-cm distance. fR=net ("background") cpm from a scanning bed measurement of 100 nc C s 137 uniformly distributed in an Alderson phantom. S2 Figure of merit: ^ ^ + { ) '

Energy S2 S2

interval S m f m fR ( 1 +f)

(keV ) (x W 3) (x 1 0 ' 3)

5 6 - 80 2 0 0 1 61 0 . 11 ' 6 6 59

5 6 -1 0 0 3 3 8 5 126 0 .2 3 91 74

5 6 -1 2 0 4 4 4 1 194 0 .3 7 1 0 2 75

5 6 -1 4 0 5242 2 6 2 0 .4 9 105 71

5 6 -1 6 0 5881 318 0 .6 0 109 6 8

5 6 -2 0 0 67 9 3 4 4 8 0. 85 103 . 56

2 8 - 1 2 0 600 3 243 0 .4 6 1 4 8 ' 1 0 2

3 2 - 1 2 0 588 7 231 0 .4 4 150 104

4 0 - 1 2 0 5543 2 1 9 0. 41 140 99

4 8 r l 2 0 504 8 208 0. 39 1 2 2 8 8

5 6 -1 2 0 4 4 4 1 194 0 .3 7 1 0 2 75

(d) Comparison of the 8 X 4 in crystal and the thin crystals. For case (a) (Table II), we can compare the two crystals with regard to their statistical performance when only small gamma contamination is present. The ratio of the optimum values is 1. 30 in favour of the thin crystals. This means that in detecting small amounts of Sr 90 in humans, the thin crystals will give J 1. 30 = 1.14 times lower standard deviation if the subject is measured for the same time with both detector assemblies, or, in order to give an equal statistical accuracy, that the thick crystal requires 30% more time; for the measurement. When the thin crystals are used, much of this time is consumed by the measurement in chair geometry necessary to determine the potassium content. Unfortunately, multichannel analyser imperfection sets the lower energy interval for measurements with the 8 X 4 in crysta l at 56 keV, so in practice the thin crystals exhibit a figure of merit 70% higher than that of the large crystal. 102 L. G. BENGTSSON

For the case of a highly contaminated subject a discussion similar to that in section 3 must replace the statistical comparison of the two detector types. .In this case the statistical errors are of less importance than other errors in the determination of .the contribution of the contamination to the content in the bremsstrahlung-assay energy interval. Particularly im­ portant are the errors resulting from the change of geometry from chair with large crystal to scanning bed with thin crystals. When higher statisti­ cal accuracy is obtained, such factors as body shape and constitution also become important. . • , .

8 . CHOICE OF GEOMETRY

8.1. Significance of activity distribution in the body

From a knowledge of other work in our department, it was clear from the beginning that the chair geometry could not compete with the scanning bed geometry for the assay of small amounts of beta emitters in the human body in presence of gamma emitters. However, an accurate correction for the contribution from gamma emitters in the low-energy band is neces­ sary. NAVERSTEN [10] has investigated the dependence of the spectral shape and sensitivity on the location of the gamma emitter in the body. This was done for chair and scanning bed geometry at the gamma energies 0.364, 0.669 and 1. 31 MeV. As expected great variations were revealed for the chair geometry. As a result of the levelling effect of the scanning bed geo­ metry the variations are smaller.. The variations of spectrum with the lo­ cation of the gamma emitter in the body must be kept small so that the cor­ rection for contributions from gamma emitters is only slightly dependent on the location of the activity in the human body. The already described severe disadvantage of the chair geometry is not balancèd by the increased detection efficiency. F o r the thin crystals, the figure of m erit fo r the chair geometry in bremsstrahlung assay for an object without gamma contami­ nations exceeds that of the scanning bed geometry by 50%, when the same measuring time is used. However, the increased sensitivity of the chair geometry impairs the statistical performance of the assay in case ofahighly contaminated subject, since the "background" increases. While there is almost no difference, between the ordinary backgrounds in the bremsstrahlung band for-chair and for scanning bed geometry, the sensitivity in the brems­ strahlung band for Csi37 is 53% higher for the chair geometry than for the scanning bed geometry. The dependence of the chair geometry sensitivity on the location of the activity iii the body is, of course, also a disadvantage in the case of bremsstrahlung produced. In our Alderson phantom; boné- equivalent 2 2 % СаС1г solution containing Sr90+Y90 was put into the three tubes shown schematically in Fig. 5 and given the arbitrary indications liver, trunk and fem ur. W ith the 8 X 4 in crystal, measurements were made in chair and scanning bed geometry. The results shown in Table IV confirm the great variations expected for the chair geometry and the levelling effect of the bèd'geometry, when averaging prone and supine positions. For this average the minimum value of the three locations is 80% of the maximum; the corresponding value for the chair geometry is 42%. ' BETA BREMSSTRAHLUNG DETECTION 103

F i g . 5 -

Sites of Sr90 bone-equivalent solution in the Alderson phantom.

TABLE IV

SENSITIVITY FOR DIFFERENT TUBES IN A HUMAN PHANTOM FILLED WITH BONE:EQUIVALENT 22% CaCl2 SOLUTION OF Sr 90+Y90

Measurements were made with the 8 X 4 in crystal. . . For tube locations see Fig.5. . ■ .. S =net counts in the en ergy band 56-160 keV per ■ _minute p er m icrocu rie Sr?0 + Y 90. ,

. s

(cpm/jjc Sr50+ Y s0), ' G eom etry Fem ur , Trunk Liver

Scanning bed

Prone 278 362 . 528

Supine 2 9 4 315 186

A verage 286 •339 3 5 7 .

C hair 493 604 1162

8.2. Variation of background with weight 1

Especially towards lower energies the background varies considerably with the weight of-the measured object. Fig. 6 shows the spectral distri­ bution of the change in photon number, measured with an 8 X 4 in crysta l, when 100 kg of sugar is introduced in'the empty chair. At 150 keV the rela­ tive change in the background counting rate is more than 20%. However, for comparison of the geometries it is sufficient to note that the variation with weight is of the same order for both geom etries. Also, the backgrounds are almost equal when no sugar is used. . 104 L. G. BENGTSSON

BA *S INDIC ATE - 1 5. D.

I +5 S ; ^ i с Ё * 3 !/ot •X I \

Г J 1 0 --— -1 0 100 200 300 400 500 600 700 800 PHOTON ENERGY . (keV)

Fig. 6 ’

Spectral distribution of change in photon number when 100 kg of sugar

is introduced in the empty chair ( 8 X4 in crystal).

For both geometries, the sugar phantom shape may be changed con­ siderably for a given weight with only small (less than 1 %) background changes, as long as the average human body shape for different weights is maintained. At the beginning a scanning bed geom etry with 35 cm between the bed and crystal surface was tried. In this geometry the background in the energy band 28-160 keV was a few per cent higher for a small compact phantom than for a tall, slender one of the same weight, when measured with the thin crystals. The variation of background with shape of object obliged us to abandon the 35-cm scanning bed geometry.

220

ç 210 £

1Л I 200 о о

190

0 10 20 30 40 50 60 70 SO 90 PHANTOM WEIGHT ( k g )

Fig. 7

Background counts in the .bremsstrahlung assay energy interval 56-160 keV for different weights of human-shaped sugar phantoms (scanning bed geometry, 8X4 in crystal). BETA BREMSSTRAHLUNG DETECTION 105

For the 45-cm scanning bed geometry the background change is pro­ portional to the weight of the subject seen by the detector. Fig. 7 shows the background variation in the energy band 56-160 keV when human shaped sugar phantoms are measured with the 8 X 4 in crystal in the 45-cm scanning bed geometry. Only small deflections from the straight line dependency do or can occur. The increase amounts to 1.7% per 10 kg increase in phantom weight. For the chair geometry the variation with weight is not so well established. In conclusion we may say that the 45-cm scanning bed geometry: a. shows little variation in sensitivity with the location of the radioactive source in the body; b. shows only small variation in background when the object's shape is changed, and; c. has a well defined dependence of background on object weight.

Therefore, the subsequent investigations concern only that geometry.

9. CALIBRATION OF THE SCANNING BED GEOMETRY FOR MEASUREMENT OF Sr^0 IN A HUMAN SUBJECT

For the thin crystals, the calibration of the scanning bed geometry was described in [8]. Y90 was injected into a human subject who was measured 1 h and 82 h after injection. In the energy interval 28-120 keV the 82 h measurement gave 202 cpm р е гц с Y 90. In section 8.1 measurements of Sr90+y90 in tubes in the Alderson phantom filled with bone-equivalent solution were described. Similar measurements were also made for the thin crystals. The scanning bed geo­ metry gave, for femur, trunk and liver, 230, 284 and 305 cpm in 28-120keV band per дс Sr90+Y90, respectively. These figures are of the same order of magnitude as obtained in the human measurement. On the average with the thin crystals in the 28-120 keV energy band, the phantom measurements give 83% of the counts obtained with the 8 X 4 in crystal in the energy inter­ val 56-160 keV. Therefore, it seems reasonable to assume that the cali­ bration factor fo r the 8 X 4 in crystal fo r Y 90 iniahumanis 202/0.83 = 224 cpm per juc. For S r90 + Y90 the figure should be somewhat higher, owing to the 0.54 MeV beta rays from Sr90. The angular dependence of the two detector assemblies in their respec­ tive optimum energy intervals was found to be roughly the same for the various angles possible during a scanning bed measurement. Accordingly, the ratio of the sensitivities from a scanning bed phantom measurement should be roughly the same as the ratio from a point source measurement with the detector fixed. The data obtained from such a measurement - giving a thin crystal sensitivity that is 86% of that of the 8 X 4 in crystal - confirm this conclusion since it is close to the 83% obtained from the phantom measurements in scanning bed geometry. 106 L. G, BENGTSSON

10. EVALUATION OF THE INTERFERENCE FROM Cs 137 AND POTASSIUM

The contribution in the bremsstrahlung energy interval from nuclides other than the pure beta emitter to be assayed is strongly dependent on the shape and the weight of the measured subject. Evaluating this dependence is one of the main tasks when large amounts of these nuclides are present and disturb the bremsstrahlung spectrum. • This task can not be accomplished with sufficient accuracy if the lo­ cation of activity in the body is markedly inhomogeneous. If, however, as is the case to a good approximation for caesium and potassium, the nuclide is evenly distributed in the body, phantom measurements can give us infor­ mation about the contribution in the bremsstrahlung energy band. The re­ sults in the following paragraphs were obtained with the 8 X 4 in crystal in scanning bed geometry. The low-energy band fraction discussed in sections 10. 1-10. 3 is the ratio of counts in the energy interval used for the brems­ strahlung measurement, 56-160 keV, and counts in the energy band used for Cs!37. determination, 568-744 keV.

10.1. Aspects of the influence of phantom weight

In order to find the low-energy band fraction for subjects having dif­ ferent weights, human-shaped phantoms of different weights were con­ structed. The first phantom type consisted of 10 kg bottles having a cross­ section of 303 cm2 and containing a solution of Cs137 in distilled water. The bottles were positioned to simulatethe human shape as much as possible. The low-energy band fraction to a good approximation increased linearly between 40 kg and 100 kg. . However, when the results were applied to measurements of human subjects, contradictory results were obtained. Obviously, the type of phantom mentioned above gave too rough an approximation to .the human. Therefore measurements of the low-energy band fractions for linear phantoms having different'cross-sectional areas were made. Each of these phantoms was constructed from bottles of equal size, filled with Cs137 in distilled water. The low-energy band fraction varied roughly linearly with the cross-section, from 0.18 for 0 cm2 to 0.62 for 303 cm2. These m easure­ ments thus confirm that bottle phantoms that are supposed to simulate human shapes must contain bottles with diameters close to the diameters of different parts of the human body. .

10.2. Calibration for. different weights , -

The experience from 10.1 could now be used to get a calibration for different weights. The bottle phantoms were definitely unsuitable for lower weights, since the cross-section would be too high whèn compared with the limbs of a light human subject. For lower weights the Alderson phantom should work well. The filled phantom contains about 50 kg solution; the skeleton weighs about 5 kg. The plastic envelope, weighing about 10 kg, should be added to this. The total weight is thus 65 kg. The weight without envelope is 55 kg. Empirically it was found that the best fit to data from human measurements was obtained if to the phantom was attributed an "equi- BETA BREMSSTRAHLUNG DETECTION 107

TABLE V

COMPARISON OF DIFFERENT PHANTOMS USED TO EVALUATE THE LOW-ENERGY BAND FRACTION FROM Csi37 IN HUMANS OF DIFFERENT WEIGHTS The phantoms were filled with Cs137 solution. , , , . .. . ., , • counts in 56-160 keV The low -en ergy band fraction is the ratio ------— ;— r „„ „ ,, .—rr counts in 568-744 keV

Low -energy cpm in W eight D iam eters Phantom type band 568-744 keV (kg) ■ ( c m 2 ) ■ fraction per nc Cs137

1 0 -kg bottles 50 303 0 .6 2 4 . 9 linear phantom

Alderson 55 . 0 .6 2 5 .3 hum an

1 0 -kg bottles 55 ■ • 3 0 3 ■ 0. 71 5 .5 human shaped (interpolated)

• 1 0 -kg bottles [ 1 ] 1 0 0 3 0 3 ' 0 .8 1 5. 1 human shaped

2 + 5 + 10-kg 1 0 0 303 0 .8 4 4 . 8

bottles [2 ] 198 human shaped . 113 valent"- weight of 55 kg. For the higher weights, the bottle phantoms would do better. For confirmation, a bottle phantom which also contained bottles having cross-sectional areas of 113 cm2 and 198 cm2 was measured. The result is-shown in Table V. A linear relationship between' the points from the Alderson phantom and from the 100-kg bottle phantom (see Table V, system (2)) should give a fairly accurate correction for the weight de­ pendence of the low-energy band contribution from Csi37 in people of aver­ age body build. However, the correction must be less accurate when applied to very thin or very stout subjects. In Fig. 8 are shown spectra from a Cs137 solution in the Alderson phantom and in the 100-kg bottle phantom (system (2)\ A spectrum from a point source measurement is also shown:. An increase below the backscatter peak at about 200 keV appears when the scattering mass increases from 55 kg to 100 kg. • ■ ' ■

10.3. Influence of spectrometer resolution -

The resolution of the 8 X 4 in crystal was deliberately changed by chang­ ing the gain of one photomultiplier. - The relative width at half maximum of the 662-keV gamma line from an uncollimated Cs137 point source at 40 cm 108 L. G. BENGTSSON

F ig . 8

Spectra of a Cs137 solution uniformly distributed in the Alderson phantom and the 100-kg bottle phantom (scanning bed geometry, 8X4 in crystal).

For comparison the spectrum of a Cs117 point source is shown; for this spectrum the scale is not applicable. distance was determined. This resolution was, for three different adjust­ ments, 10.3%, 11.5% and 13.1%. F o r each resolution a 100-kg bottle phantom containing Csl37 solution was measured in the scanning bed geometry. No significant (less than ± 1%) difference in the low-energy band fraction was found for these resolutions, over a wide range. The resolutions in the measured phantom spectra were 10.8%, 12.2% and 13.9%.

10.4. Potassium

For the correction for the potassium contribution in the low-energy band the same considerations as in 10. 2 are valid. For assay of potassium the energy interval 1280-1560 keV was used. The low -energy band fractions from this interval are shown in Table VI. Figure 9 shows spectra from potassium in the Alderson phantom and in the 100-kg bottle phantom (see Table VI, system (1)). Here, the scattering increases over a wider part of the spectrum than for Cs137.

11. SYNTHESIS OF SPECTRA

The aim of this paper was primarily to evaluate a method for assay of pure beta emitters in presence of large amounts of gamma emitters. In 6-3, it was demonstrated that for this purpose the 8 X 4 in crystal would be the best detector. This and the following section contain a description of the use of this detector for human measurements and some results. In principle, the measured spectrum of a subject should be compared with a spectrum synthesized from a sugar background spectrum and the spectra due to the measured radionuclide content of the subject. The dif­ ference should be due to other nuclides than those giving distinct peaks in the spectrum. И the subject is known to contain a certain radionuclide, the BETA BREMSSTRAHLUNG DETECTION 109

TABLE VI

COMPARISON OF DIFFERENT PHANTOMS USED TO EVALUATE THE LOW-ENERGY BAND FRACTION FROM POTASSIUM IN HUMANS OF DIFFERENT WEIGHTS

The phantoms w ere filled with a solution of KC1 p. a. F o r bottle phantoms only 10-kg bottles with a cross-sectional area 303 cm2 were used.

, л. , í • x, x- counts in 56-160 keV The low-energy band fraction is the ratio cou¿ts in i 280-Í560 keV

The Cs137 band fraction is the ratio cou^ s in 568 744 keV counts in 1280-1560 keV

cpm in W eight Low -energy Cs137 band Phantom type 1280-1560 keV (kg) band fraction fraction per g К

Alderson human 55 0 . 7 4 0 .2 5 0 .4 0 ’

1 0 -kg bottles 55 0 .8 0 - - human shaped (interpolated)

1 0 -kg bottles [ 1 ] 1 0 0 0. 94 0 .2 7 0 .3 8 human shaped

Spectra of a solution of KC1(A.R. ) uniformly distributed in the Alderson phantom and the 100-kg bottle phantom. The ordinate gives counting rate per unit weight '

of potassium (scanning bed geometry, 8 X4 in crystal). amount of this nuclide can be assayed; if nothing is known about the cause of an observed excess of counts in a spectrum, the maximum pos­ sible amounts of any interesting nuclide, such as Sr90) can be established. 110 L. G. BENGTSSON

In practice, synthesis of whole spectra would be possible with electronic data processing. For manual treatment only three energy bands are used at present, 56-160 keV for bremsstrahlung assay, the Cs137 energy band 568-744 keV, and the potassium energy band 1280-1560 keV. Background counting rate and contributions'from the net counting rates in the 568-744-keV and 1280-1560-keV bands are subtracted from the measured counting rate in the 56-160-keV band. Of course some extension of the number of energy intervals could be made if other distinct peaks were found in the subject's spectrum. .

12. RESULTS - .

12.1. Measurements • ■ •

The results from measurements of six subjects are given in Table VII. The predicted standard'deviation (SD) in the rest in the bremsstrahlung assay energy band would correspond to about 20 nc S r90 fo r a 38. 4-min m easure­ ment of a subject (Eq.(l), section 7. For five cases the measured rest is less than 3 SD, corresponding to 60 nc Sr90 . In the sixth case (IT) is present a rest that can not be due to the statistics alone. Since this subject had worked with patients given Co57 (photon energy 122 keV) it is probable that she had a slight contamination (< 5 nc) of this nuclide.

12.2. Important sources of error ,

Since we work with the low-energy end of the spectra from large objects, we easily get severe disturbances from many sources. In paragraph 12.1 it was pointed out that it is difficult to make corrections for a contamination of a subject with small amounts of radionuclides, (cf. also 7a.) In the case quoted the contamination was probably caused by occupational use of radionuclides. Another source of contamination is fall-out. RUNDO and NEWTON [11] found, in April 1962, about 0.3 nc Zr95+Nb95 in average English subjects. 1 nc Z r 95 + Nb95 would give a contribution in the bremsstrahlung band c o rre ­ sponding to about 20 nc S r90. We must not forget the fall-out nuclide Sr9o which is now present in average European subjects, in quantities of a few nanocuries. Thus we have also a small contribution from this nuclide. The contribution in the low-energy band from radionuclides naturally occuring in the body, besides K40, should be negligible. The most important nuclide is Ra2260f which we have about 0.1 nc; This will give a contribution in the low-energy band corresponding to a few nanocuries of Sr90. There is also a small probability that short-lived radon decay products may ac­ cumulate to the order of magnitude of 0.1 nc in the subject's lungs. We have experience that paper pyjamas exposed to the open air may become contaminated with non-negligible amounts of air-borne activity. T h erefore, the paper pyjamas used by the subjects in Table VII were measured before measurement of the subject. This should always be done, and the pyjamas should be stored with no access to circulating air. ■ Fig. 10 shows the rest spectrum from a measurement of a subject probably wearing TABLE VII

SYNTHESIS OF SPECTRA FROM SIX SUBJECTS The first five subjects'were measured 19.2 min in prone and 19.2 min in supine position in the scanning bed geometry, by the 8 X 4 in crystal. Backgrounds were obtained from a varying number of 19.2-min measurements of 50-kg sugar phantoms before and after the subject measurements. The subtracted background has been corrected according to the subject's weight. Subject AM was measured 39. 0 +39. 0 min. M = male, F = fem ale. EA RMSRHUG DETECTION BREMSSTRAHLUNG BETA

Rest in 56-160 keV Net count in 38. 4 m in P re d icte d d o u b le SD 5 6 - 1 6 0 5 6 8 * 7 4 4 1 2 8 0 - 1 5 6 0 contributions from S u b je ct S ex W eight in the rest keV keV keV C s 137 and К

(S r 90 (n u m b er (S r 90 * equivalent, (kg) • of counts) ‘ equivalent n c) n c)

JL M 96 8-380 8 2 7 8 2 0 4 2 . + 1 2 6 + 13 4 0

BM M ‘ 76 8 3 0 0 9 3 6 9 . 2 3 6 9 - 2 2 - 2 ■ 4 4

IL M 7 2 . 8 6 7 2 9 7 1 4 2 1 4 5 ■ + 4 6 1 ' + 4 9 4 4 .

IT F 70 4 6 9 4 4 4 4 2 1 7 6 5 + 6 7 2 + 7 2 ' 3 5

GB M 68 5 3 9 9 5 9 1 6 2 0 9 6 + 3 4 . + 4 4 0

AM F .. 50 2 6 0 6 2 8 7 4 1 4 9 0 + 33 + 4 2 3

B ack ground . 50 « 8 6 0 0 « 4 4 0 0 . « 3 2 0 0

* 1 0 n cC s137 corresponds to about 2 0 0 0 counts (correction for potassium must be made).

* * 100 g К corresponds to about 1500 counts. 112 L. G. BENGTSSON contaminated paper pyjamas. The shape of the rest spectrum is very similar to that of the bremsstrahlung spectrum shown for comparison. When this subject was re-measured in new pyjamas after two weeks no significant rest was found (Table VII, subject AM). One has to be careful also, in the case of cotton overalls. Washing sent to the laundry has probably become contaminated with Zr95+Nb95 in a few cases [12] . In order to remove contamination in the hair and on the skin, all sub­ jects should take a shower before the measurement. They should leave jewelry, watches and spectacles in the changing room. Especially in the measurements of subjects with small amounts of gamma contamination, it is very important to keep.the background of the whole-body counter stable. The whole-body counter at Lund has a good ventilation, but even then it was found that background variations appeared, causing errors in the assay corresponding to more than a hundred nano­ curies S r9o. The background is stable if the temperature of the counter is kept constant and if the door to the whole-body counter is always closed, except for the necessary passage in and out. Background measurements must be run before and after the measurement of the subject in order to establish the stability. TABLE VIII

MINIMUM DETECTABLE AMOUNTS (MDA) OF Sr90 IN HUMAN SUBJECTS Scanning bed geometry, 8 X 4 in crystal. The energy interval 56-160 keV is used for the bremsstrahlung assay. A subject weighing 60 kg with a potassium content of 120 g is assumed. A = Only purely statistical errors are considered. MDA corresponds to 3 SD in the rest in the bremsstrahlung assay energy interval. В = To the statistical error is added the contribution from a 1% SD in the determination of the low-energy band fraction, t = Total time for measurement of subject and background (min).

MDA (n c S r9») C s 137 content A В (n c) t = 2 0 t = 4 0 t = 1 2 0 t = 2 0 t = 40 t = 1 2 0

0 103 . 73 42 103 73 42

50 1 2 1 85 49 123 87 53

1 0 0 . 137 94 53 143 103 6 8

2 0 0 164 1 1 0 63 182 138 105

500 2 2 6 142 82 3 42 254 226

1 0 0 0 3 0 2 183 105 517 459 433

12.3. Reliability of the spectrum synthesis procedure

Minimum detectable amounts (MDA) with the above method for the esti­ mation of the Sr9o content of the human body are given in Table VIII. Be­ BETA BREMSSTRAHLUNG DETECTION 113 sides the purely statistical errors, we must account for the errors in the determination of the low-energy band fraction. Non-uniform activity distri­ bution, varying body build and calibration errors may cause uncertainty in the proper value of the low-energy band fraction. We assume that the accuracy in our determination of the low-energy band fraction for Cs137 is such that the relative SD is 1%. In column B, Table VIII, are given the mini­ mum detectable amounts of Sr9° when the mentioned e rro r adds to the purely statistical one. The low-energy band fraction error is considered as a random error and the MDA corresponds to three times the total SD in the assay. Data are calculated for a total measurement time of .20, 40 and 120 min. They were adjusted to the practical situation so that the back­ ground is measured at least for 5 min before and 5 min after the subject measurement. When this condition is fulfilled, the time is shared on subject and background measurements so that the best statistical accuracy is obtained. The results in Table VIII indicate that for contaminations of less than 100 nc Csi37 only the purely statistical errors need be considered, so the minimum detectable amount of a pure beta emitter is that which corresponds to 3 SD in the rest in the bremsstrahlung assay energy band. This is not contradicted by the results from measurements of human subjects in Table VII. ,

F i g .10

Rest spectrum from a subject measurement. The rest is probably due.to contamination of the paper pyjamas worn; The rest corresponds to about 200 nc Sr90.

For comparison the solid line shows a Sr90 spectrum from a phantom measurement, (scanning bed geometry, 8x4 in crystal.) 114 L. G, BENGTSSON

13. CONCLUSIONS

From Table VIII we find that the minimum detectable amount of S r90 with the above method is of the order of 50 nc. The present contamination levels of Sr90 in the average European and American population amount to a few nanocuries. The method can thus at present be used only to state the maximum possible burden. The most conservative interpretation of the ICRP recommendations gives a maximum permissible body burden (in bone) of 70 nc ЭгЭО^ which is of the same order of magnitude as the mini­ mum detectable amount. F o r individuals subject to occupational exposure to ionizing radiation, the maximum permissible burden (in bone) is 2000 nc. Thus the method couM'be'used for a direct assay of occupationally acquired contamination, even in the presence of large amounts of gamma emitters.

ACKNOWLEDGEMENTS

Dr. Kurt Lidén has contributed substantially to this paper. I am deeply indebted to him for his sustained interest, advice and encouragement during the work. The International Atom ic Energy Agency supplied the point brem s­ strahlung source. Grants from the Division of Radiological Health, Bureau of State Services, Public Health Service, USA and the Swedish Atomic Re­ search Council are gratefully acknowledged.

REFERENCES

[1] • LIDÉN, K. , Proc. 2nd UN Int.. Conf. PUAE 23 (1958) 133. [2] MEHL, H. G. , Strahlentherapie 102 (1957) 569. [3] MATS, C. W. , TAYSUM, D. H. , FISHER, W. and GLAD, B. W. , Hlth Phys. 1 (1958) 282. [4] KELLERSHOHN, O. , HERSZBERG, B. and MARTIN, J. , Strahlentherapie,' Sonderband 38 (1958) 331. [5] TUB IANA, М ., ALBAREDE, P. and NAHUM, H. , Proc. 2nd UN Int. Conf. PUAE 26 (1958) 217.

[ 6 ] HEIDELBERG, Joan G. , STROTA, P.G. , DEWEY, W.C. and ROSE, R.G. , Amer. Í. Roentgenol. 90 (1 9 6 3 ) 3 2 5 . [7] WINCHELL, H.S. , Lawrence Radiation Laboratory Rpt UCRL-9755 (1961) unpublished.

[ 8 ] LIDÉN, K.V.H . and McCALL, R .C ., "Low-energy photon detectors in whole-body counting", IAEA, Vienna (1962) 145. [9] GOLDMAN, M. , YOUNG, L. and EDMONDS, R. , Rpt. UCD-108 University of California (1963) 90. [10] NAVERSTEN, Y. , A two-crystal scanning-bed counter for accurate determination of the whole-body activity, these Proceedings. [11] RUNDO, J. and NEWTON, D. , Nature, Lond. 195 (1962) 851-853. [12] BENGTSSON, L. G. , NAVERSTEN, Y. and SVENSSON, K .G ., these Proceedings.

■ . DISCUSSION-

J. RUNDO (Chairman): Was any difference observed in the scattered radiation emitted from the bodies of persons óf different size, sim ilar to the difference found with the two phantoms? If so, how are corrections made for this? ■ G. BENGTSSON: Corrections were largely based on empirical data from measurements performed on human subjects (weighing 50-95 kg), to supplement the phantom measurements. . AN IMPROVED CHEST PHANTOM FOR STUDIES OF PLUTONIUM AND AMERICIUM IN HUMAN LUNGS

R.G. SPEIGHT, C.O , PEABODY AND D. RAMSDEN . RADIOLOGICAL AND SAFETY DIVISION, ATOMIC ENERGY ESTABLISHMENT WINFRITH, DORCHESTER, DORSET, ENGLAND

Abstract — Résumé — Аннотация — Resumen

AN IMPROVED CHEST PHANTOM FOR STUDIES OF PLUTONIUM AND AMERICIUM IN HUMAN LUNGS. Work is in progress on the problem of measuring plutonium and americium isotopes, either separately or as mixtures, in the human lung. They may be present as point or distributed sources and are detected directly by their low energy X- and gamma-rays. The high absorption of the X- and gamma-rays in the soft tissue, ribs, sternum, spine, e tc ., makes it difficult to correlate the response of external detectors with a configuration of internal sources. A semi-empirical method must therefore be used in which sources are placed inside a realistic phantom having similar dimensions and absorption properties to the human chest. The design and construction of such a phantom are described and its method of use illustrated. Experimental results are given for the photon absorption of the constructional materials at energies down to 9.25 keV. Their degree of equi­ valence to human tissue is discussed.

NOUVEAU MODELE DE THORAX ARTIFICIEL POUR LES ÉTUDES SUR LA CHARGE DE PLUTONIUM ET D'AMÉRICIUM DANS LES POUMONS. Les-auteurs étudient actuellement comment doser le plutonium et Г américium - soit isolés, soit mélangés - dans les poùmons. Ces radioisotopes peuvent être présents sous forme de sources ponctuelles ou répartis dans un certain volume, et on les décèle directement à l’aide des rayons X et gamma de basse énergie émis. En raison de l'absorption d'une grande quantité de rayons X et gamma par les tissus mous, les côtes, le sternum, la colonne vertébrale, e tc., il est difficile d’établir une corrélation entre la réponse de détecteurs externes et la configuration de sources internes. Il faut donc recourir à une méthode semi-empirique qui consiste à placer les sources à l'intérieur d'un thorax artificiel dont les dimensions et les caractéristiques d’absorption sont semblables à celles du thorax humain. Les auteurs indiquent les caractéristiques et décrivent le montage d'un fantôme de ce genre et en expliquent le mode d'utilisation. Ils présentent les résultats d'expériences sur l'absorption des photons par les matériaux de construction, pour des énergies descendant jusqu’à 9,25 keV. Ils examinent dans.quelle mesure ces matériaux sont équivalents à des tissus humains.

УСОВЕРШЕНСТВОВАННЫЙ ФАНТОМ ГРУДНОЙ КЛЕТКИ ДЛЯ ИССЛЕДОВАНИЯ СО­ ДЕРЖАНИЯ ПЛУТОНИЯ И АМЕРИЦИЯ В ЛЕГКИХ ЧЕЛОВЕКА. Ведутся работы для оешения проблемы измерения содержания изотопов плутония и америция (как отдельно, так и их смеси) в легких человека. Изотопы могут находиться в виде точечных или распределенных ис­ точников, они определяются прямым способом по их рентгеновскому и гамма-излучению низ­ кой энергии. Высокое поглощение рентгеновских и гамма-лучей мягкими тканями, ребрами, грудиной, позвоночником и т .д . затрудняет расчет реакции внешних детекторов на данную

конфигурацию внутреннего источника. Следовательно, н у ж н о использовать полуэмпирический метод для которого источники помещаются внутри реального фантома, имеющего те же из­

мерения и т у же степень поглощения, что и грудная клетка человека. Описываются конструк­ ция и изготовление такого фантома и иллюстрируется метод его использования. Приводятся экспериментальные результаты относительно поглощения фотонов строительными материалами при энергиях ниже 9,25 кэв. Обсуждается степень их эквивалентности человеческим тканям.

SIMULADOR TORÁCICO PERFECCIONADO PARA ESTUDIAR EL PLUTONIO У AMERICIO EN EL PULMON HUMANO. La memoria describe los estudios que se están realizando para medir los isótopos del plutonio y del americio en el pulmón humano, sea por separado, sea como mezclas. Dichos isótopos pueden constituir fuentes puntiformes o difusas, y se detectan directamente por sus rayos X de baja energía o sus rayos y . Los tejidos blandos, las costillas, el esternón, la espina dorsal, e tc., absorben gran parte de los rayos X y y, por

115 1 ie R. G. SPEIGHT e t al.

lo que es difícil relacionar la respuesta de un detector externo con la configuración de la fuente interna. Esto obliga a emplear un método semiempirico en que las fuentes se colocan en un simulador de dimensiones y propriedades absorbentes similares a las del tórax humano. La memoria describe el diseño y la construcción del simulador y explica cómo.se emplea. Se dan resultados experimentales para la absorción de fotones por los materiales con que está construido, operando con energías tan bajas como 9,25 keV. Se discute su grado de equivalencia con los tejidos humanos. .

1. INTRODUCTION

Many people are engaged in work where plutonium and americium are used in various forms and proportions. In spite of the careful precautions which are taken there is always a possibility that these materials may es­ cape from their immediate containment and be subsequently ingested or in­ haled. It is felt that their most likely route of accidental intake at Winfrith is by inhalation as relatively insoluble particles. Work is therefore in pro­ gress on the methods of measuring such particle activity deposited in the respiratory system, using external radiation detectors. The chest phantom described here was built as part of this work. It is hoped that these methods will eventually be used to supplement and improve on the present methods of assessing body content by analysis of excreta. Much of the present day plutonium is extracted chemically from uranium fuel elements which have been subjected to high total neutron irradiation. Under these conditions the extracted plutonium contains plutonium-239 plus appreciable quantities of plutonium-240 and plutonium-241, produced by successive neutron capture during irradiation. The plutonium-241 decays with a 13.3-yr half-life to americium-241 so that the latter can build up to a considerable extent in a few months after the plutonium extraction. The plutonium may be determined by measuring its associated 13.6, 17.4 and 20.4 keV uranium L X-rays while americium-241 may be measured by means of its 60-keV gamma rays in addition to its L X-rays.* These low energy photons may be measured either by means of scintillation de­ tectors using thin crystals of sodium iodide or by proportional counters. LIDEN [1] has reported on the use of scintillation detectors and TAYLOR and RUNDO [2] have described work with proportional counters. Both types of detector have been examined at Winfrith and are being further developed.

2. THE NEED FOR A REALISTIC CHEST PHANTOM:

The basic performance characteristics of suitable photon detectors are already fairly well known when used to measure point sources of plutonium or americium behind uniform absorbers in simple physical and geometrical conditions. The conditions in the case of the human respiratory system are, however, considerably more complicated because the plutonium or americium may be present as few or many particles anywhere in the system. The ma­ terial between the particles and the detector may consist of many combi­ nations of tissue, body fluids, air space or bone, producing considerable

Note: for convenience the uranium L X-rays are frequently called "plutonium 17-keV X-rays. - LUNG PHANTOM FOR PLUTONIUM AND AMERICIUM 117

attenuation at these low photon energies. Thus the response of a given de­ tector to a given total amount of internal radioactive material is virtually impossible to predict theoretically because of this anatomical complexity. Conversely it is difficult at present to see a method by which a given de­ tector indication can be used to calculate the amount of internal radioactivity. The problem is particularly difficult when only plutonium is present, but in many practical cases it may be possible to measure the associated americium-241 by means of its more penetrating 60-keV gamma ray, thus using it as a.built-in tracer for the plutonium. Adequate investigations and experiments with human subjects are ob­ viously impracticable. It was therefore decided to construct a phantom of the same shape and size as a typical human chest and having sim ilar ab­ sorption characteristics for low energy photons. This phantom is now being used in an attempt to solve some of these problems on a semi-empirical basis, by inserting known sources at known internal positions whilst studying the response of external detectors.

3. POSSIBLE TYPES OF PHANTOM

3.1. Laminated phantoms

Several authors, including COFIELD [3], have described the use of chest phantoms constructed by stacking sheets of material such as untempered Masonite. Cofield describes a phantom fifteen inches square by nine inches high with some central air spaces to simulate the lower density of the lungs. We constructed a similar phantom from nineteen sheets of Mix-D tissue equivalent material (which is discussed in section 4.1).- The sheets were 30 cm by 30 cm by 1 cm thick. The bottom sheet was unperforated but the next sixteen sheets were perforated by an array of 2.5-cm diam. holes. The number and positions of the holes were varied from sheet to sheet in such a way that, when assembled, two regions of reduced bulk density were pro­ duced in the phantom. The patterns of the holes were chosen so as to make these regions of the same general size, position and average density ás those of the lungs in a human chest. Slots were cut in the eighteenth sheet in the f.orm of a plan view of the sternum and ribs. A bone equivalent mixture was made from wet plaster of Paris and calcium phosphate (see section 4. 2) and poured into these slots, thus simulating the ribs and sternum. A top sheet of unperforated Mix-D was then added to complete the phantom. Standard medical radiographs were taken of the phantom and showed approximate equivalence to a 25-cm thick human chest. It was soon found that although this simple phantom is useful in the case of higher energy gamma-ràys (for example 184-keV uranium-235 gamma rays) it is not sufficiently realistic to be of use at 17 and 60 keV.

3.2. Anthropomorphic phantoms

Several types of sophisticated and detailed anthropomorphic phantoms are available, such as those manufactured by Alderson Research Laboratories, 118 R. G. SPEIGHT e t al.

Inc., in the USA. None of these were however available to us at the start of the work and it was decided to make one at Winfrith to meet our particular requirem ents. A basic specification’ for the required phantom was:

1. a thin, tissue equivalent, optically transparent outer case moulded as an accurate replica of a human chest; 2. a thoracic cage from a human skeleton, enclosing artificial lungs made of tissue equivalent material to the same size, shape and den­ sity as those in an average human being; 3. provision for the addition of other organs as required; 4. all other tissue spaces to be filled with a bulk tissue equivalent ma­ terial, preferably non-liquid for convenience in handling; 5. provision for rapid dismantling of the phantom so as to modify its internal arrangement or to insert sources where necessary; 6. simplicity of construction, consistent with adequate realism, and economical manufacture from easily available materials; 7. radioactive impurities in materials to be negligible.

4. POSSIBLE CONSTRUCTIONAL MATERIALS

4. 1. Tissue equivalent materials .

Tissue is a term which covers a wide range of body materials with a corresponding variation in X-ray and gamma ray scattering and absorption characteristics. The tissue may consist of fat or muscle and may contain varying amounts of body fluids such as blood and water. For most ex­ perimental work'it is not possible to use living tissue. Samples of post mortem animal tissue are also of doubtful value because of the loss of fluid. Many attempts have therefore been made in the past to produce synthetic materials equivalent to an "average" type of tissue. SPIERS [4] has shown that the absorption and scattering effects are governed by the electron density n (usually expressed as electrons per cm3) and the effective atomic number Z of the material. The photoelectric ab­ sorption at low photon energies is critically dependent on Z, whilst n chiefly determines the Compton scattering at higher energies. He gives formulae whereby n and 2 may be calculated from the composition and density of the material. Tissue equivalence is obtained by adjusting the composition until n and Z are equal to those for tissue. Spiers gives a value of n equal to 3.36X 1023 electrons per cm3 and Z equal to 7.42 and 7.46 fo r biceps m uscles. In contrast his values for subcutaneous fat are 3.17X 1023 and 5.92. ROSSI and FAILLA [5] consider that soft tissue can be represented chemically by (C 5H40 O1 8 N)x, for which n = 3.31X 1023 and Z = 7.19. LINDSAY and STERN [6]; calculate that for "average human tissue" n = 3.32X 1023 and Z = 7.33. Some of the materials of interest in this application are discussed below and values of n and Z are shown in Table I. Water has n= 3.36X 1023 and Z = 7.42 and is therefore closely equivalent, to tissue at medium to high photon energies but may show slight differences from average tissue at low energies. . LUNG PHANTOM FOR PLUTONIUM AND AMERICIUM 119

JONES and RAINE [7] developed a solid material which they called "M ix-D " for which n = 3.36X 1023 and Z = 7.47 thus making its properties more similar to those of water than to average tissue at low energies. Its composition is paraffin wax 60.8%, polyethylene 30.4%, magnesium oxide 6.4%, titanium dioxide 2.4% and it is supplied by Messrs. Astor, Boissellier and Lawrence of West Drayton, Middlesex. It can be cast, cut or machined but is somewhat brittle. LINDSAY and STERN [6] devised a tissue-like material which they called "Lincolnshire Bolus". It consists of small pellets made from a mix­ ture of sucrose or glucose and "magnesium carbonate levis" (approximate formula 3 MgC03. Mg(OH)2. ЗН2О). The pellets contain 87% by weight of sugar and 13% by weight of magnesium carbonate levis. They are nearly spherical in form with a mean diameter of 2.3 mm. The bolus is supplied by Boots Ltd., Nottingham. The material is used in bulk form consisting of the closely packed pellets separated by air interstices and has an effective bulk-n of 3.34 X 1023 . Z is stated by the authors to be w ell within 2.0% of that for water (7.42) and within 1.0% of that for "average human tissue" (7.33). It must therefore be in the range 7.3 to 7.4. STACEY, BEVAN and DICKENS [8] developed a material named "Temex" which is based on fluid natural rubber and is commercially available from James Girdler and Co., London. Its composition is adjusted so that n equals 3.32X 1023 for a density of 1.01 and Z is very nearly equal to that of tissue and water. Tem ex can be moulded in any desired shape with internal cavities or bones. It can also be supplied in a foamed form to a specified density which is obviously ideal for construction of phantom lungs. ; "Perspex" (polymethylmethacrylate plastic) is a useful constructional material with n = 3.84X 1023 and Z = 6.47. It is therefore only roughly equi­ valent to water and average tissue but is a suitable material for use in small thicknesses, as for the outer shell of a phantom. . Polyvinyl chloride is available in a foamed form known as "Polyfoam", obtainable commercially from Leslies (Polyfoam Division), London. This' has a bulk density of only 0.045 g/cm3 and gives comparatively low attenu­ ation at low photon energies. This attenuation can be increased by trapping suitable materials in the air cavities. We carried out some experiments in which Polyfoam was made to absorb a warm aqueous solution of Agar. When allowed to cool for some time the Agar set to a gel form inside the pores of the material, thus trapping an amount,of water greater than that which would normally have been retained by,the Polyfoam. As the water slowly evaporated from the foam the bulk density gradually decreased. The density could be stabilized for twelve to twenty-four hours at any selected value by coating the foam with a plastic 'skin obtained by painting on a thick solution of Perspex in chloroform. This was convenient for short experi­ ments as simulated lung material but the skin was not sufficiently im­ permeable for the Agar-loaded.foam to be of permanent use. Some typical results are given in section 5.

4. 2. Bone and equivalent m a terials

In general the natural material can be used with the marrow space im ­ pregnated, if necessary, with tissue equivalent material. This has not been 120 R. G. SPEIGHT e t al.

TA BLE I

VALUES OF n AND Z FOR VARIOUS MATERIALS

n Z M a te ria l R eferen ce (e le ctro n s (E ffe c tiv e per c m 3) atomic number)

Muscle (biceps) Spiers [4 ] . З .З б х Ю 23 7.42 and 7.46

Subcutaneous fat Spiers [4] . 3 . 1 7 Х Ю 23 5 . 92

Soft tissue Rossi and Failla [5] 3 . 3 1 Х 1 0 23 7 . 1 9

Average human tissue Lindsay and Stern [ 6 ] 3 . 3 2 X 1 0 23 7 . 3 3

W ater Spiers [4] З .З б х Ю 23 7 . 4 2

M ix -D Jones and Raine [7] З .З б х Ю 23 7 . 4 7

Lincolnshire Bolus Iindsay and Stern [ 6 ] 3 . 3 4 Х Ю 23 7 . 3 to 7 . 4

T e m e x Stacey et al. [ 8 ] 3 . 3 2 X 1023 Close to tissue

P erspex C a lc u la te d 3 . 8 4 X 1 0 23 6 . 4 7 '

necessary for our present low energy work. For some experimental pur­ poses we have used a mixture of plaster of Paris (CaS03. 2H20) and calcium phosphate (Ca3(P04)2) to give similar predicted absorptive properties to those of the mineral component of bone.

5. EXPERIMENTAL MEASUREMENTS ON CONSTRUCTIONAL MATERIALS '

5. 1. Photon attenuation

The most suitable m aterials appeared to be foamed Temex for the lungs, Perspex for the moulded outer case in Lincolnshire Bolus as a "dry fluid" for fillin g in tissue spaces. Inform ationis available in the references quoted on their absorptive properties at photon energies down to about 30 keV (e f­ fective) using X-ray machines but not in the 10-to20-keV region. Measure­ ments were therefore carried out for these and several other materials in this region and also at 60 kèV (for comparison with results in the literature). Three sources of radiation were used, americium-241 giving 60-keV gamma rays, plutonium (free from americium) giving X-rays at 13.6, 17.4 and 20.5 keV and germanium-71 giving 9.25-keV gallium X-rays. Measure­ ments at 60 keV and in the 17-keV region were carried out by means of a detector incorporating a sodium iodide crystal of 12-cm diam. and 0.2 cm thick. Counting rates were determined in the 37-to 81-keV and 8-to 30-keV energy bands respectively. The sources were placed on the axis of the de­ tector and 10 cm away from the surface of the crystal. Samples of the ma- LUNG PHANTOM FOR PLUTONIUM AND AMERICIUM 121 terials 5 cm square and of various thicknesses, were placed against the sources, between them and the detector. The measurement geometry was intermediate between narrow and broad beam and was representative of that which applies for a source of small dimensions separated by several centi­ metres of body t'issue from a large detector. A proportional counter was used for the 9.25-keV measurements. In this case the source was positioned 2 cm away from the 0.001-in thick aluminium window of the counter in a collimator of 30 degrees total angle. Here the samples were of 1-cm diam. and of thicknesses up to 1 cm. The observed transmissions for the three energy regions are plotted on a logarithmic scale against thickness in Figs. 1, 2 and 3. The observed points at 9.25 keV (F ig . 1) and 60 keV (F ig . 3) fa ll on ess en tia lly straight lines but those for the plutonium source follow curves (Fig. 2) because of the different absorptions at the three component energies. In comparing materials the initial part of these lines and curves was used to give the ex­ perimental first half-thickness values. These are given with their cor­ responding linear absorption coefficients in Table II.' The statistical ac­ curacy of the observed counting rates was limited to 5 to 10% by the rela­ tively small plutonium and americium sources available. The error on the values in Table II is about 10%.

1.0 \

0.5 ¡N 4 \ X \ 4

IT t \ \ \ \ ! \ О \ 4? \ «л \ \ z \ i X V \ z< о \x \ « 0.05 \ a Í \ ? \ \' л \ \ v\ \ Л \ w \ ТЕМ EX (X) \ к \wUE («)\ MIX 'D' WAX {O \ 0.01 4 0.1 0 2 Û3 0.4 0 5 0.6 0.7 0.0 Q9 1.0 1.1 1.2 13 1.4 1.5 THICKNESS (cm)

Fig-1

Transmission of 9.25-keV Ge71 К-capture X-rays

The materials of particular interest are Temex, Lincolnshire Bolus, water and Mix-D. Their behaviour in general follows that observed or pre- 2 RG. PIH e al. a et SPEIGHT . R.G 122

'В.

v - ‘

i * . BEEFSTE/ K ( e ) < • C'a s®S' Ч \ ч (o ) — N LINCOLNSHIRE , BOL JS ( î>) ' .4 'XЧ 4 N TE МЕХ (X WATER (e FRACTIONAL TRANSMISSION

0 1 2 3 4 5 6 7 THICKNESS (cm) THICKNESS (cm)

F ig . 2 F i g - 3

Transmission of 17-keV Pu239 X -ra y s Transmission of 60-keV Am 241 gamma rays LUNG PHANTOM FOR PLUTONIUM AND AMERICIUM 123

TA BLE II

MEASURED FIRST HALF THICKNESS IN VARIOUS MATERIALS

Corresponding F irst h alf lin ear Energy ' M a te ria l thickn ess absorption coefficient (k eV ) (c m ) ( c m -1)

9 . 2 5 0 . 1 4 5 . 0

W ater 17 0 . 6 0 1 . 2

60 4 . 9 0 . 1 4

- 9 . 2 5 0 . 1 3 5 . 3

M ix-D w ax 17 0 . 4 9 1 . 4

60 . 5 . 4 0 . 1 3

17 0 . 4 8 1 . 4 Lincolnshire Bolus 60 4 . 9 0 . 14

9 . 2 5 0 . 1 8 3 . 8

Tem ex rubber 17 0 . 5 5 1 . 3 '

60 4 . 9 0 . 1 4

17 0 . 4 8 1 . 4 Lean beefsteak

60 ~ 8 ~ 0 . 1

Rigid foamed polystyrene 17 ~ 2 5 ~ 0 . 0 3

Polyfoam 17 ~ 1 6 ~ 0 . 04

Agar loaded Polyfoam (0. 25 g/cm î) 17 2 . 7 0 . 2 6 .

Agar loaded Polyfoam (0. 34 g/cm 3) 17 0 . 8 7 0 . 80

Perspex acrylic sheet 17 0 . 6 6 1 . 1

Bone substitute 17 ~ 0 . 1 [ C a S 0 4 ,2НгО + С аз(Р04)г to 2% P]

Note: The 17-keV measurements were done with a plutonium source. •

dieted in the references quoted. At 60 keV their half-thickness values in our geometry lie within the range 4.9 to 5.4 cm so that, within experimental error, their performances are virtually identical at this energy. For plu­ tonium the half-thickness values for Temex, Lincolnshire Bolus, and Mix-D fa ll within a range of 0.48 to 0.55 cm compared with'water at 0.60 cm. The 124 R.'G. SPEIGHT e t al. agreement between water and M ix-D is close at 9.25 keV with half-thicknesses of 0.14 and 0.13 cm whereas the half thickness fo r Tem ex is 0.18 cm. No measurements were made on Lincolnshire Bolus at this low energy because the mean diameter of the pellets is about two half thicknesses. Irregularities in the bulk packing density of the pellets would have been too great at the small sample thicknesses involved to give meaningful absorption figures. In general our prelim inary results show that the behaviour of these four materials is very similar at energies down to about 13 keV but that Temex shows less attenuation than water or Mix-D at 9.25 keV. We have at present no means of determining which of these materials shows the closest equi­ valence to tissue in the 10-to 20-keV range although the calculations of STACEY et a l. [ 8 ] appear to indicate that Tem ex might be the closest equivalent in this range. The results of some measurements with plutonium and americium for beef steak are shown in F ig s . 2 and 3. The steak was re frig e ra te d and sliced so as to make convenient, firm samples. Its photon absorption in small thicknesses for plutonium was similar to that of the tissue equivalent materials but became less than theirs at great thicknesses. At these greater thicknesses most of the 13.6-keV X-rays have already been absorbed leaving a relatively larger proportion of the 17.4-and 20.5-keV X-rays. Thus the divergence is greater at these effectively larger energies, and is even more pronounced at 60 keV. This behaviour is consistent with a good match for Z and a poor match for n between the steak and the other m aterials because of the loss of blood and tissue fluid. The plutonium curve for Perspex (Fig. 2) shows the expected divergence due to a mismatch in both Z and n. It is however suitable for use as the thin outer shell' of the phantom. . One approximate measurement was made on live flesh with plutonium. The flesh used was the muscular region between the thumb and first finger of a fully extended spread hand. The average thickness was about 3 cm. The observed point in Fig. 2 appears to fall on the curve for steak but the direction of the probable errors in the measurement are such that it should probably fall nearer to the tissue equivalent material curves. A suitable sample of foamed Temex was not available for these measure­ ments but the measurements on the solid material indicate that the half thickness at the required lung density (0.27.. g/стЗ) should be about 2.1 cm. The two curves for Agar-loaded Polyfoam show that this half thickness is also in the range of density (0.25 to 0.34 g/стЗ) which is easily obtained for this material. Thus phantom lungs could easily be fabricated from Poly­ foam for short experiments, although they would not be suitable for per­ manent use. The plutonium curve for foamed polystyrene shows that its attenuation is so low that it may be used to simulate air spaces in the phantom or as structural supports where the addition of any appreciable absorption would be undesirable.

5. 2. R ad ioactivity in m a teria ls

Samples of Temex, Lincolnshire Bolus, Perspex, Polyfoam, Agar, foamed polystyrene, Mix-D, bone and bone substitute were checked for the LUNG PHANTOM FOR PLUTONIUM AND AMERICIUM 125 presence of any undesirable radioactive impurity. None was detectable in the gamma and low energy X -r a y regions.

6 . CONSTRUCTION OF THE PHANTOM

The outer case is fabricated from Perspex sheet of 1.5-mm thickness as shown in Figs. 4 and 5. An accurate plaster cast was made of the chest of one of the authors (R. G. Speight) and from this four Fibreglass moulds were made, two of the front half and two of the rear half. These moulds were made by laying sheets of glass fibre mat on to the greased surface of the cast, applying a suitable resin and catalyst, then allowing to dry and set. A sheet of Perspex was placed over one of the front moulds and heated by several large gas flames until it softened. This process was continued until the softened sheet had assumed a shape near to that of the mould. The second front mould was then warmed and pressed down on the Perspex thus forming it into the shape of the front of the chest. A back section was pro­ duced in a sim ilar way. They were then fitted with flanges at sides, shoulders, waist, neck and armpits. Perspex plates were attached by means of nylon screw s at the neck, arm pits and waist.

' Exploded diagram of Perspex shell

A human thoracic cage with clavicles and scapulae was obtained from Adam Rouilly and Co. , London. It was supplied with vertebrae and polythene intervertebral discs mounted on a Perspex rod. The ribs, clavicles and 126 R. G. SPEIGHT e t al.

F ig . 5

Partially assembled chest phantom scapulae were secured and articulated by nylon thread. The perspex rod was attached by nylon screws to the base and neck plates of the case. The bones were not impregnated with tissue equivalent material since this is not necessary for the present low energy work. The manufacturers supplied a pair of phantom lungs moulded from foamed Temex to the required density of 0.27 g/cm3. The size and external shape are those of an average pair of lungs whilst the interior has air ca­ vities corresponding to the air passages of the 'main bronchus. The. lungs were mounted inside the thoracic cage by nylon threads. The trachea and upper part of the bronchi were represented by polyvinyl chloride tubes of the appropriate diameters. The upper end of the trachea was fastened to the neck plate and the lower ends of the bronchial tubes were fixed into the upper bronchial passages of the lungs. The oesophagus was represented by a long polythene tube secured to the top and bottom plates. Finally 45 lb of Lincolnshire Bolus were carefully poured in through one of the armpit holes with the phantom on its side, thus filling all the other phantom cavities. The Lincolnshire Bolus slowly settles with time and LUNG PHANTOM FOR PLUTONIUM AND AMERICIUM 127

further small amounts can be added through a hole in the top plate to restore the level. ' No metal components were used in the phantom.

7. FIRST EXPERIMENTS WITH THE PHANTOM

7.1. Radiographs

A normal parallel beam radiographie exposure was made of the chest phantom and a human subject (R.G. Speight). The general appearance and range of photographic densities were very similar in each case, apart from the fact that the phantom skeleton is of Indo-Asian origin and its bones are 'therefore slightly smaller than those in an average European subject. An- attempt was made to take a radiograph with a radioactive source inside and with films arranged as a cylinder around the phantom. It was found however that very little detail was obtainable partly because of the unavailability of sufficiently intense low energy sources and also because of the unsuitable geometry. It was however possible to distinguish obvious shadows cast by the sternum, spine, etc. i.e. where little response would - be obtained on an external detector for an internal low energy photon source.

7. 2. Internsd plutonium source

After an accident involving the inhalation of insoluble particles many of the smaller ones will be quickly distributed throughout the lungs via the bronchi and bronchioles. During the period immediately after inhalation it is, however, likely that appreciable quantities of any larger particles will be found in the region of the upper bronchi. Also at a much later time it is possible for particles to be removed from the lung by phagocytic action via the lymphatic system. These particles will accumulate in the lymph nodes surrounding the upper bronchi and lower trachea. Thus K R E Y and others [9] have reported the presence of long standing burdens of plutonium in the lymph nodes. These are therefore two possible cases in which inhalation of insoluble particles containing plutonium could produce appreciable con­ centrations of activity in the tracheo-bronchial region. These cases were roughly simulated by lowering a 0.5-g plutonium source down the phantom's trachea to the point where this joins the left and right bronchi. An external detector was used to plot contours of the X-ray intensity at the surface of the phantom. The detector contained a thin sodium iodide crystal with a 1.3-cm diam. collimator and a photomultiplier feeding into a,single channel analyser covering the 15-to 19-keV region. The de­ tector was held against the phantom surface and normal to it at many po­ sitions defined by a grid system drawn on the phantom; This grid system will be used for defining positions and recording counting rates on the wide variety of chest shapes and sizes which will be encountered. It is shown in F ig . 6 and consists basically of parallel beam projections of two rec­ tangular grid planes on to the front and back of the phantom. The results recorded on the grid were used to obtain counting rate contours. Figs. 7 128 R. G. SPEIGHT e t al.

Projection of reference planes

Intensity contour (front) with in tern ai p lu to n iu m sou rce LUNG PHANTOM FOR PLUTONIUM AND AMERICIUM 129

F i g . 8 . ,

Intensity contour (back) with internal plutonium source

and 8 show these contours transferred to the phantom. These make it very obvious that, in this case, even a large detector placed over the main part of the lung would give negligible counting rates compared with those for a much smaller one above the sternal notch.

8 . CONCLUSIONS

A description has been given of the development of a phantom for studies of plutonium and americium in human lungs. Work has started on^the problem of deciding the optimum size, characteristics and positions of ex­ ternal detectors by semi-empirical methods. The typical experiment mentioned in the previous section illustrates that the best choice of detector may not always be a large, complex one in the conventional lung position but may be a smaller, simpler one elsewhere. In future experiments plu­ tonium and americium sources will therefore be placed in many positions in the respiratory system either by insertion down the trachea or in small ca­ vities cut into the lungs. Response contours will then be obtained for se­ v e ra l sizes and types of detector. Additional source entry tubes and additional organs will be added to the phantom as required. For instance sources will be distributed in a simulated trácheo-bronchial lymph node system in order to develop the optimum method for detection of the long standing burdens which may occur in this region. 130 R. G. SPEIGHT e t al.

ACKNOWLEDGEMENTS

We have received generous help from our collea:gues at Winfrith throughout this work. We would like to thank Dr. A.M . Laylee and Sister G.G. Walker of the Medical Branch for much advice and technical help, par­ ticularly in the manufacture of the plaster cast of the chest. The radiographic work was-undertaken by M rs. P. Martin and Mr. R.L. Grant. Mr. J.R. Hazell, Mr. E. A. Sherwood and their workshop staff successfully developed the techniques for moulding the Perspex shell. Mr. F.H. Passant tested the constructional materials for the absence of radio-activity. We also wish to thank Mr. A. J. Stacey (Royal Marsden Hospital) and Mr. B.M. Wheeler (James G irdler and Co. ) who designed and produced the foamed Tem ex lungs.

REFERENCES

[1] LIDEN, K.V.H . and McCALL, R .C ., "Low-energy-photon detectors for whole-body counting", Whole- ‘ body counting, IAEA, Vienna (1962) 145. [2] TAYLOR, В. T. and RUNDO, J ., A progress report on the measurement of plutonium in vivo, Rpt. AERE/ .R -4 1 5 5 (1 9 6 2 ) 12 pp. [3] COFIELD, R. E., Health Physics 2 (1960) 269-87: In vivo gamma counting method of determining uranium lung burden in humans, Oak Ridge, T e rm ., Rpt Y -1 2 5 0 (1 9 5 9 ) 33 pp. [4] SPIERS, F. W., Brit. J. Radiol. 19 (1946) 52. ' . [5] ROSSI, H.H. and FAILLA, G ., Nucleonics 14 2 (1956) 32.

[ 6 ] LINDSAY, D. D. and STERN, B.E., Radiology 60 3 (1953) 355. ' [7] JONES, D.E.A. and RAINE, H .C., Brit. J. Radiol. 22 (1949) 549.

[ 8 ] STACEY, A.J., BEVAN, A.R. and DICKENS, C.W., Brit. J. Radiol. 34 (1961) 510. ' . [9] KREY, P.W .. BOGEN, D. and FRENCH, E ., Nature, Lond. (1962) 4838. NOUVEAU COMPTEUR PROPORTIONNEL DESTINE A LA DETECTION IN VIVO DE TRACES DE PLUTONIUM DANS LES POUMONS

. A. LANSIART ET J.-P.MORUCCI . COMMISSARIAT A L'ÉNERGIE ATOMIQUE, SACLAY, FRANCE

Abstract — Résumé — Аннотация — Resumen

NEW PROPORTIONAL COUNTER SUITED TO THE IN VIVO DETECTION OF PLUTONIUM TRACES IN THE LUNGS. The authors explain the construction of a gas proportional counter with an effective detecting surface o f 250 cm 2 and a sensitivity approximately equal to that obtainable with a scintillating crystal. In simplified terms, it consists of two superposed multi-wire counters within the same chamber, one serving as detector and the other to reduce background by anti-coincidence. The window is of beryllium

1 mm thick and the instrument is filled with xenon at a pressure of 2 a tm . An important feature of this counter is the excellent uniformity of detection over the entire window surface. This uniformity is obtained by means of corrections for the edge effects, these corrections being very easy to apply. The wires аге25дт diam. and 3 cm apart, so that it is possible to use much lower voltages than are customary in this type of detector and to suppress parasitic pulses due to spurious emissions. The authors discuss the data obtained from a contaminated animal.

NOUVEAU COMPTEUR PROPORTIONNEL DESTINE A LA DETECTION IN VIVO DE TRACES DE PLUTONI­ UM DANS LES POUMONS. Les auteurs ont entrepris la réalisation d‘un compteur proportionnel à gaz dont l'efficacité de détection est voisine de celle que l'on peut obtenir avec un cristal scintillateur et dont la surface utile de détection est de 250 cm 2. Schématiquement, il comprend deux compteurs multifile superposés dans la même enceinte: le premier sert à la détection, le second à la réduction du mouvement propre par anticoincidence. La fenêtre est en béryllium de 1 mm d'épaisseur. Le gaz de remplissage est du xénon sous la pression de 2 atm. Une des caractéristiques importantes de ce compteur est l'excellente homogénéité de détection sur toute la surface de la fenêtre. Cette homogénéité est obtenue grâce à des corrections des effets de bords, très simples à mettre en œuvre. Les fils employés ont 25 ¡i de diamètre et sont distants de 3 cm . Dans ces conditions, il est possible d'utiliser des tensions plus basses que celles couramment utilisées dans ce type de détecteur et de supprimer les causes d'impulsions parasites dues aux microclaquages. Les résultats obtenus sur une contamination animale sont discutés.

НОВЫЙ ПРОПОРЦИОНАЛЬНЫЙ СЧЕТЧИК ДЛЯ ОБНАРУЖЕНИЯ СЛЕДОВ ПЛУТОНИЯ В ЛЕГКИХ IN VIVO. Автооами создан газовый поопооииональный счетчик с эффективностью обнаружения, близкой к той, которую можно получить с кристаллическим сцинтиллятором, причем полезная площадь обнаружения его равна 250 см2. . Он состоит из двух многонитевых счетчиков в одном и том же корпусе: ^первый исполь­ зуется для обнаружения,, а второй — для восстановления собственного движения лутем анти­ совпадения. Окно закрыто бериллиевой пластинкой толщиной 1 мм. Газовым заполнителем

корпуса является ксенон под давлением 2 атмосферы. - ‘ Одной из важных особенностей этого счетчика является замечательная гомогенность обнаружения на всей поверхности окна. Эта гомогенность достигается благодаря весьма простым и доступным поправкам краевых эффектов. Используемые нити имеют 25 микрон в диаметре, и отстоят друг от друга на Зсм . В этих условиях становится возможным применять более низкие напряжения, чем те. которые обычно используются в настоящее время в счетчике такого типа, и устранить причины появ­ ления посторонних импульсов, возникающих в результате микропробоев. Обсуждаются результаты, полученные на зараженных животных.

131 132 A. LANSIART et J.-P . MORUCCI

NUEVO CONTADOR PROPORCIONAL ADAPTADO A LA DETECCIÓN IN VIVO DE TRAZAS DE PLUTONIO EN LOS PULMONES. Los autores han emprendido la construcción de un contador proporcional de gas, cuya eficacia de detección es aproximadamente igual a la que se puede obtener con un cristal de centelleo cuya superficie útil de detección sea de 250 era2. Esquemáticamente, comprende dos contadores de conductores múltiples superpuestos dentro del mismo recinto; el primero sirve para la detección y el segundo para la reducción de la actividad de fondo por antico­ incidencia. La ventana es de berilio, de 1 mm de espesor. El gas de relleno es el xenón, a una presión de 2 atmósferas. Una de las características importantes de este contador es la excelente homogeneidad de detección en toda la superficie de la ventana. Esta homogeneidad se logra gracias a correcciones de los efectos de los bordes, muy sencillas de efectuar. Los alambres empleados tienen 25 jj de diámetro y un espaciamiento de 3 cm . En estas condiciones, es posible utilizar tensiones más bajas que las corrientemente empleadas en este tipo de detector y suprimir las causas de los impulsos parásitos debidas a descamas locales. Se examinan los resultados obtenidos en un caso de contaminación animal.

INTRODUCTION

Nous présentons les résultats du premier prototype monté au labora­ toire et dont les caractéristiques sont: a) diam ètre utile: 13 cm, b) nature du gaz: xénon - méthane (90% - 10%) et c) pression: 850 mm Hg. Un compteur de 18 cm de diamètre utile, prototype d'une petite série d' appareils, est en cours de réalisation, mais les essais de tenue en pres­ sion ne sont pas encore terminés. Ce compteur a été spécialement étudié en vue de pouvoir déceler une contamination pulmonaire par le plutonium 239. Nous avions déjà réalisé des détecteurs à scintillation, utilisant des cristaux minces de Nal(Tl) de 3 cm de diamètre et dont le mouvement propre est de 1 cpm dans la bande d'énergie correspondant au rayonnement XLas- socié au rayonnement a du plutonium. Mais il est malaisé d'utiliser des photomultiplicateurs de grand diamètre avec des cristaux minces en raison de leur mauvaise uniformité de sensibilité; par ailleurs, le nombre de photoélectrons reçus par scintillation est insuffisant pour séparer les dif­ féren tes ra ies X L. Pour ces deux raisons, nous avons décidé la réalisation d'un compteur proportionnel de grand diamètre. Les résultats présentés par KIEFER et MÂUSHART [1] à Vienne en 1961, nous parurent encourageants (voir aussi [2]). . Nous nous sommes orientés vers un compteur avec fenêtre en béryllium (à cause de sa transparence aux rayons X du plutonium) permettant de ré­ aliser, un appareil étanche pouvant être rempli d'un gaz coûteux tel que le xénon éventuellement sous pression. r ' Dans ces conditions nous pouvions être assurés d'un bon rendement de détection. Il était de plus séduisant d 'u tiliser un compteur multifile plans parallèles comme compteur proportionnel. En effet le volume utile de dé­ tection et le volume contribuant effectivement à donner des implusions de compteur peuvent être rendus voisins. De plus, la géométrie de l'appareil se prête au montage de deux comp­ teurs superposés dont 1 ' un peut servir de blindage à 1 ' autre en étant monté en anticoïncidence. ' NOUVEAU COMPTEUR PROPORTIONNEL 133

DESCRIPTION

Une coupe schématique du compteur est donnée à la figure 1. Le détecteur est équipé d'un four à calcium destiné à l'absorption de l'oxygène dégazé par les matériaux du compteur. Il1 est muni d'une fenêtre en béryllium de 1 mm d'épaisseur, ce qui permet de faire le vide, afin d'ef­ fectuer le remplissage du xénon-méthane sous pression. L'anode est constituée.par une couronne en cuivre sur laquelle sont tendus des fils de molybdène recuit, dont le diamètre (25 ц ) est garanti à 2% près, et qui sont distants de 26 mm. De part et d'autre de cette grille de comptage, deux autres couronnes équipées de fils de 150 ц tous les 5 mm servent de cathode et les fils de masse et les fils HT est de 12 mm. Mais l'inconvénient d'un tel dispositif réside dans ce qu'on appelle «l'e ffe t de bord», c'est-à-dire les distorsions du champ sur les supports des grilles de comptage. On constate en effet que, dans un compteur mülti- fils, l'amplitude de l'impulsion électrique donnée par une source 7 ou X collimatée décroît du centre au bord le long d'un fil ou transversalement aux fils . Sachant que, dans un compteur cylindrique, pour une tension donnée, on augmente la hauteur de l'impulsion en diminuant le diamètre de cathode, on a obtenu un effet correcteu r des effets de bord en augmentant l'ép aisseu r de la couronne de masse, pour une distance entre les réseaux de fils de masse et de HT constante. • De plus, les compteurs actuellement connus ont un nombre de fils haute tension important, par exemple un écartement de 1 cm pour un diamètre de 20 cm et des fils de 50 ц . En agissant sur 1'effet correcteur, il a été possible de diminuer d'un facteur 3 le nombre de fils et ceci sans perte de rendement ni d'uniformité de réponse. La haute tension pour une même amplitude de signal est ainsi diminuée de 25% et atteint une valeur proche de celle qu' on obtiendrait avec un seul fil. On peut en outre utiliser des fils plus fins, ce qui permet de diminuer encore la haute tension. '

RÉSULTATS

Le spectre d'une source de 239Pu étendue dont le diamètre est supérieur à celui de la fenêtre de détection est donné à la figure 2. La haute tension utilisée est de 1600 V. On constate que les résolutions pour les raies X Lde 13,6 keV et 17,2 keV sont de 15% et 13,3% pour le xénon-m éthane sous 850 m m H g. Dans les m êm es conditions, e lle s sont de 11,4% et 10,9% pour 1 'argon-m éthane. Mais ceci était attendu. En effet, des données spectroscopiques et pho­ tochimiques du méthane, on constate que sa photodécomposition présente un spectre d' absorption entre 1450 et 850 Â. Mais si les photons UV émis lors de la désexcitation de l'argon sont fortement absorbés en dissociant le méthane, il n'est pas de même de tous les photons émis par le xénon et ces photons au cours de l'avalanche créent une ionisation supplémentaire qui détériore la résolution [3]. 134 A. LANSIART et J.-P. MORUCCI

• Figure 1

Coupe schématique du compteur.

Figu re 2

Spectre d'une source étendue de 239Pu.

Gaz Xe + CH4: 850 mm Hg. Fenêtre en béryllium; épaisseur: 1 mm. Anticoïncidence interne. Bruit de fond déduit. ,

10 15 ÉNERGIE (keV) NOUVEAU COMPTEUR PROPORTIONNEL 135

Toutefois, les résolutions que nous avons obtenues avec le xénon- méthane sont meilleures que celles obtenues couramment dans les comp­ teurs de conception classique remplis de xénon-méthane, l'ionisation sup­ plémentaire créée par des photons UY tombant sur la cathode constituée par des fils étant négligeable.

EFFICACITÉ DU COMPTEUR

L'efficacité du compteur dépend évidemment de l'énergie du rayonne­ ment incident et la figure 3 donne la courbe de rendement théorique (nombre de photons X détectés par rapport au nombre de photons X arrivant normale­ ment) du détecteur pour un rayonnement le traversant normalement, l'épais­ seur du'gaz étant de 24 mm. Les rendements sont respectivement de 63%, 44% et 32% pour les raies de 13,6, 17,2 et 20,2 keV pour du xénon sous 1 atm et de 82%, 6 8 % et 54% pour.du xénon sous 2 atm. Dans le cas du détecteur correspondant à la version industrielle, où l'épaisseur utile de gaz est alors de 32 mm, les rendements théoriques attendus sont pour 13,6 keV, 17,2 keV, 20,2 k eV resp ectivem en t de 70%, 55%, 41% sous 1 atm de xénon et 85%, 78% et 63% sous 2 atm. Expérimentalement, on trouve des rendements supérieurs aux rende­ ments théoriques du fait que le gaz présente un parcours supérieur à 24 mm pour des rayons X traversant obliquement le compteur. . Par exemple, avec une source étendue de diamètre 15 cm à 45 mm de la fenêtre de notre compteur, les rendements expérimentaux sont respec­ tivem ent pour les ra ies X ( de 13,6, 17,2 et 20,2 keV: 69%, 5.1%, 37%, ce qui correspond à un parcours moyen de 30 mm. Mais ces rendements pour

ÉNERGIE (keV)

Figu re 3

Efficacité théorique d'un compteur proportionnel rempli de xénon sous 1 atm et 2 atm

et muni d'une fenêtre en béryllium de 1 m m .

Epaisseur de gaz traversée: 2,4 cm. 136 A. LANSIART et J.-P. MORUCCI une source de dimension donnée, sont fonction de la distance de la source au détecteur, la distance moyenne traversée par le flux de rayonnement X dépend de l'angle solide de'détection. Ils diminuent quand la distance aug­ mente. ' D'autre part, pour une contamination profonde, l'absorption par les tissus va être plus grande pour les photons pártant obliquement. Il est donc nécessaire que le rendement pour les photons pénétrant normalement à la fenêtre soit le plus élevé possible. Il est fort utile de pouvoir déterminer la profondeur de la contamination. Or les différentes raies Xj, ne subissent pas la même absorption spécifique dans lés tissus.-. Du rapport entre le nombre de raies X, reçues pour les différentes énergies, il est possible de déterminer l'épaisseur d'ion ab­ sorbant connu. ' 1 ■ 1 Si le détecteur n'a pas le même rendement de détection pour chaque raie Xj en fonction de la distance, l'étalonnage devient très pénible. ' La figure 4 montre des résultats expérimentaux obtenus avec le xénon- méthane et 850 mm Hg de pression. Pour une même activité due à la raie 13,6 keV, on voit comment va rie l'a c tiv ité due aux autres raies. . On constate que pour une source ponctuelle entre 5 et 10 cm de la fenêtre, les rendements varient de 4% pour le 20,2 keV et 2,5% pour le

Figu re 4 ■

Résultats expérimentaux obtenus par le xénon-méthane.

Gaz xénon-méthane: 850 mm Hg. Fenêtre en béryllium; épaisseur: 1 mm.

Source ponctuelle de 239Pu.

------: source à 1 0 cm de la fenêtre...... : sou rce co n tre la fe n ê tre . NOUVEAU COMPTEUR PROPORTIONNEL 137

17,2 keV; entre 0 et 5 cm de la fenêtre, les rendements varient de 15% pour le 20,2 keV et de 12% pour le 17,2 keV.

MESURES D'ABSORPTION. APPLICATION A LA DÉTECTION D'UNE CONTAMINATION DE PLUTONIUM ET A LA MESURE DE PROFONDEUR

Nous avons interposé entre une source de 15 cm de diam ètre, de 0,2 ц с . de 239Pu et le compteur, différentes épaisseurs de plexiglas entre 5 mm et

Figure 5

Courbes d'absorption..

Gaz xénon-méthane: 850 mm Hg. Fenêtre en béryllium: épaisseur: 1 mm.

Source étendue de 2S9Pu. Bruit de fond déduit. . .

(T 4 sans absorbant. .

(T 5 mm de plexiglas.

( з 4 1 0 mm de plexiglas.

(7 15 mm de plexiglas.

( Г 2 0 mm de plexiglas.

25 mm de plexiglas.

(T4 30 mm de plexiglas. 138 A. LANSIART et J.-P. MORUCCI

I 30 mm pour simuler une absorption, la distance source-fenetre du compteur étant 45 mm. Les courbes dont données à la figure 5 en coordonnées semi- logarithmiques.

ÉPAISSEUR DE PLEXIGLAS (mm)

Figure 6

Absorption totale et absorption des 3 pics du 239Pu en fonction de l'épaisseur de plexiglas traversée (ordonnées de gauche) '

■ surface pic 17 keV surface pic 17,2 keV apports surjaC£ pjc jg g kgy e surface p¡c 2 0 , 2 keV

en fonction de l'épaisseur de plexiglas traversée (ordonnées de droite)-

Nous avons indiqué à la figure 6 : a) les absorptions relatives des.dif­ férentes raies et l 'absorption de la somme de 3 raies en foriction de l'épaisseur de plexiglas traversée et b) les rapports (surface 17,2 keV/ surface Î3,6 keV) et (surface du 17,2 keV/surface du 20,2 keV) en fonction de l'épaisseur traversée, les surfaces étant définies par des comptages dans des bandes centrées respectivement sur 17,2 keV, 13,6 et 20,2 keV. Ces rapports suivent une loi sensiblement linéaire et il est ainsi aisé d'ob­ tenir une estimation de la profondeur moyenne d'une contamination.

BRUIT DE FOND

Les,performances actuelles sont figurées dans le tableáu I. Ce genre de détecteur est également parfaitement adapté à la détection des raies XL des transplutoniens et la figure 7 présente le spectre d'une source de 252Cf. NOUVEAU COMPTEUR PROPORTIONNEL 139

Figure 'î

Spectre d'une source de 252Cf.

Gaz xénon-méthane: 850 mm Hg. Fenêtre de béryllium; épaisseur: 1 mm. Bruit de fond déduit.

TABLEAU I

PERFORMANCES ACTUELLES

Sans blindage A vec 15 cm de Pb Bruit de fond entre sans AC v av e c AC sans AC av ec AC 1 0 ,5 et 2 2 ,5 keV 1 2 0 c /m in 25 c /m in 60 c /m in 8 c /m in

AC = Anticoïncidence .

Nous avons obtenu pour les 3 raies XLles énergies de 14,;9 keV, 19 keV et 22 keV et les intensités relatives de 1, 1,3 et 0,5 définies à l'aide des rendements expérimentaux.

CONCLUSION

Il est possible de déceler avec ce type de détecteur une contamination sous 25 mm de plexiglas et d'en déterminer la profondeur à 3 mm près pour 140 A. LANSIART et J.-P. MORUCCI

a) une mesure d'une heure tant pour la source que pour le mouvement propre b) une source de 0,02 цс de 15 cm de diamètre. Cette erreur sera évidemment encore plus faible avec un compteur re­ pli sous 2 atm. de xénon-méthane.

REFERENCES

[1] KIEFER, H. and MAUSHART, R ., Determination of plutonium-239 body burden using gamma spec­ trometry with proportional counters, Whole-body counting, IAEA, Vienna (1962 ) 289. . [2] FESSLÉR, H ., KIEFER, H. und MAUSHART, R ., Atompraxis 7 11 (1961). [3] LANSIART, A. et MORUCCI, J.-P ., J. Phys. Radium 23 (1962) 102A. PERFORMANCE OF AN ARRANGEMENT OF SEVERAL LARGE-AREA PROPORTIONAL COUNTERS FOR THE ASSESSMENT OF Pu239 LUNG BURDENS

R. EHRET, H. KIEFER, R. MAUSHART AND G. MOHRLE . ' KARLSRUHE NUCLEAR RESEARCH CENTRE, KARLSRUHE, FEDERAL REPUBLIC OF GERMANY

Abstract — Résumé — Аннотация — Resumen

PERFORMANCE OF AN ARRANGEMENT OF SEVERAL LARGE- AREA PROPORTIONAL COUNTERS FOR ASSESSMENT OF PLUTONIUM-239 LUNG BURDENS. If plutonium-239 has been deposited in the human lung, some of the emitted soft X-rays can be measured from the outside by proportional counters. Efficiency of such measurements depends on sensitive area and geometrical arrangement of the counters in regard to the body, on absorption of the X-rays inside the counter, on reduction of the radiation in the wall or window of the counter, on the background of the counter and on stability of this background in spite of 3- and y- radiation coming from other nuclides than plutonium-239 in the body. Most of these parameters are optimized by using three large-area, thin-window, argon*filled proportional counters mounted inside a large steel-room opposite the chest and both lung lobes of a sitting patient. These counters are anti-coincidence shielded in such a way that only soft X-rays causing the origination of photo­ electrons inside the measuring counter proper are detected. The effect is enforced by passing the electrical pulses through a one-channel pulse height analyser. -Construction and performance of such an arrangement are described in the paper.

ENSEMBLE DE PLUSIEURS COMPTEURS PROPORTIONNELS A GRANDE SURFACE, POUR LA DÉTERMI­ NATION DE LA CHARGE DE PLUTONIUM-239 DANS LES POUMONS. Lorsque le plutonium 239 s'est déposé dans les poumons, on peut mesurer du dehors, à l'aide de compteurs proportionnels, une partie des rayons X mous émis par ce radionucléide. L'efficacité de ces mesures dépend de la surface sensible des compteurs et de leur disposition géométrique par rapport au corps, de l’absorption des rayons X à l’intérieur du compteur, de la réduction du rayonnement dans la paroi ou la fenêtre du compteur, du mouvement propre du compteur et de la stabilité de ce mouvement propre malgré le rayonnement Б et les rayons y émis par les nucléides, autres que le plutonium-239 contenus dans l'organisme. ' ' Pour la plupart de ces^paramètres, on procède à leur optimisation en utilisant trois compteurs proportionnels à grande surface et à fenêtre mince, remplis d'argon et montés dans une grande chambre en acier face à la poitrine et aux lobes des poumons d'un sujet en position assise. Ces compteurs sont à anti-coïncidence et protégés de façon que l'on ne puisse détecter que les rayons X mous donnant lieu à la formation de photo­ électrons à l'intérieur du compteur proprement dit. On peut intensifier cet effet en faisant passer les impulsions électriques par un analyseur d'amplitude â un seul canal; Le mémoire décrit la construction et le fonctionnement d’un ènsemble de ce genre.

ХАРАКТЕРИСТИКА РАЗМЕЩЕНИЯ РАЗЛИЧНЫХ КРУПНЫХ ПРОПОРЦИОНАЛЬНЫХ СЧЕТЧИКОВ ДЛЯ ОЦЕНКИ СОДЕРЖАНИЯ ПЛУТОНИЯ-239 В ЛЕГКИХ. При отложении плутония-239 в легких человека некоторая доля слабоинтенсивного рентгеновского излучения может быть измерена снаружи с помощью пропорциональных счетчиков. Эффективность таких измерений зависит от области чувствительности и геометрического, расположения счет­ чиков по отношению к телу, поглощения рентгеновских лучей в самом счетчике, уменьшения излучения в стенке или окне счетчика, фона счетчика и стабильности этого фона в зависи­ мости от бета- и гамма-излучения других изотопов в организме (кроме плутония-239). Большинство из этих параметров могут быть улучшены путем использования трех круп­ ных наполненных аргоном пропорциональных счетчиков с узким окошком, помещенных в боль­ шой экранированной сталью комнате против грудной клетки и обоих легких сидящего пациента. Эти счетчики являются счетчиками антисовпадений и экранированы таким образом, что об­ наруживают лишь малоинтенсивные рентгеновы лучи, вызывающие образование фотоэлектронов

141 142 R. EHRET et al.

в соответствующем измеряющем счетчике. Эффект усиливается за счет прохождения элек­ трических импульсов через один канал амплитудного анлизатора импульсов. В докладе дается описание конструкции и характеристик т акого размещения.

RENDIMIENTO DE UN DISPOSITIVO DE VARIOS CONTADORES PROPORCIONALES DE GRAN SUPERFICIE SENSIBLE, UTILIZADO PARA EVALUAR LA CARGA PULMONAR DE PLUTONIO-239. Cuando el plutonio-239 se deposita en el pulmón humano, parte de los rayos X blandos emitidos puede medirse desde el exterior, con la ayuda de contadores proporcionales. La precisión de esas medidas depende de la superficie sensible y de la disposición geométrica de los contadores con respecto al cuerpo, de la absorción de rayos X en el interior del contador, de la atenuación de la intensidad de las radiaciones en la pared o ventana del contador, de la actividad de fondo de éste y de la estabilidad de dicha actividad, pese a las radiaciones beta y gamma pro­ venientes de otros radionúclidos distintos del plutonio-239 contenidos en el organismo. El valor óptimo de la mayor parte de estos parámetros puede obtenerse utilizando 3 contadores pro­ porcionales, de gran superficie sensible provistos de ventana delgada, rellenos de argón y montados en el interior de una gran cámara de acero frente al tórax y a ios lóbulos pulmonares de un paciente que permanece sentado. Estos contadores están provistos de un blindaje por anticoinciaencias, de modo que sólo se cuenten los rayos X blandos que dan origen a fotoelectrones en el interior del contador donde se efectúa la medición. El efecto se intensifica haciendo pasar los impulsos eléctricos por un analizador monocanal de amplitud de im pulsos. Se describen la construcción de dicho dispositivo y los resultados con él obtenidos.

1. INTRODUCTION

In view of the inaccuracy of assessing the body burden of plutonium by measuring body excretions it is necessary, in the case of incorporation of plutonium after an accident, to make use of all available measured results. In the first instance, this includes an in vivo assessment of the plutonium amount in the lungs by measuring the soft X -rays in the 14- to 20-keV energy range. Various scientists have suggested and used proportional counting arrangements (for a survey of methods see [1 ] and [2 ]), but to date no equip­ ment has been found satisfactory in every respect. •' The design of a suitable detecting arrangement is a two-fold problem: increase in detection sensitivity by increasing the counter efficiency as well as by reducing background influence; optimization of size and arrangement of counters relative to the human lungs. . Various methods can be applied to solve this problem. To increase the counter efficiency the proportional counter may be run on a counting gas of high atomic number, e. g. krypton or xenon, at high pressure. Moreover, by increasing its dimensions it may be designed for a more favourable geometry (large area) as well as for high absorption (great thickness). A large counter has also another advantage: the count-rate be­ comes less sensitive to the geometrical variations of the plutonium distri­ bution inside the lung than a small counter. The reasonable increase in area is lim ited by the geom etrical dimensions of the source to be monitored, in this case the human chest. An increase in thickness is stopped as soon as a considerable fraction of the radiation, i. e. , 80-90%, is absorbed. How­ ever, all measures directed towards an increase in counter efficiency are only effective if, at the same time, the background is not considerably in­ creased. On the contrary, all possible means have to be taken to reduce back­ ground counts. These include shielding, anti-coincidence counters in con­ ASSESSMENT OF Pu233 LUNG BURDENS 143 nection with suitable shaping of the detector, energy discrimination and se­ lection of constructing materials free of own.radioactivity. For various reasons we did not want to use a counting gas of higher atomic number than argon, or run the counter above atmospheric pressure. If the entrance area for radiation is to reach some dm2 in size, the window fo il in a pressu rized counter has to be so thick, that the gain due to better absorption in the gas is again connected with an excessive absorption loss in the counter wall. Moreover, if one wants to use thin windows, counters, having a larger area, can only be operated by the flow principle because of unavoidable leakage and impurities. Counting gases of high atomic number (practically only xenon may be used; for a discussion of gases, see [3] ) have the disadvantage of being expensive. For this reason, the gas must be purified in the circulation equipment which requires additional equipment. On the other hand, for reasons of effective background reduction by using anti-coincidence.circuits, we thought it very important to have a flat design of the whole detector. By using a flat design one can have a very simple technical solution of the principle of full-space anti-coincidence [4] by arranging the counting wires as grids of a proper flatness in a common housing,one on top of the other; the counting area, then, may be of any size. The optimum dimensions of the counting area, as w ell as the most favourable arrangement of one or more counters around the human chest, can be determined only by calibration measurements. In that process the use of any phantom will constitute a source of errors, the extent of which can only be estimated. Thus, to get the most realistic results we did some calibration measurements on a human corpse.

2. DESIGN, DIMENSIONS, AND PERFORMANCE OF PROPORTIONAL

COUNTERS USED AS DETECTORS

Large-area proportional counters have been used for in vivo detections of plutonium for several years [4, 5, 6] . As an improvement, we now use a triple large-area proportional Counting system run by the flow principle on 90% argon and 10% methane as the counting gas. The dimension of the window, having a surface weight of 0.9 mg/cm2, is 1,5X30 cm2. Each gas layer of the two outer anti-coincidence counting chambers effective for ab­ sorption is^l5 mm, that of the intermediate measuring counting chamber is 40 mm. Although the anti-coincidence counting volume between source and measuring counting area contributes to a certain efficiency loss by attenuat­ ing the plutonium X-rays (some 15%) this effect is more than set off, first by the corresponding reduction in background amounting to the factor of 6-7 in the steel-chamber (see Table I), secondly by the increase in discrim i­ nating possibilities against other types of radiation because, in this arrange­ ment, external (3-radiation is not counted at all. This may become extremely important if, after an accident, the persons to be monitored have been con­ taminated, with other radioactive emitters besides plutonium. Figure 1 shows the technical set-up of the counter. In the energy band used the background without outside shielding is 19 cpm, in a body counter steel-room 15 cm thick it is less than 1 cpm. A sample of 1 /uc of Pu239 4 R ERT t al. et R. EHRET 144

T A B L E I

- INFLUENCE OF SHIELDING AND, ANTI-C.OINCIDENCE COUNTERS ON BACKGROUND COUNT-RATE

Inside of 15-cm steel-room Outside of steel-room

. 1 0 .£ - 2 2 . 5 keV . ' - 3 - 3 0 keV ‘ . 10.5-22. 5 keV 3 - 3 0 keV

facto r fa c to r fa cto r fa c to r o f b a c k ­ of b a c k ­ o f b a c k ­ o f b a c k ­ с pm cp m cp m cp m ground . ground ground ground red u ctio n red u ctio n ■ reduction re d u ctio n

No a n ti- ' ,

coincidence 92 2 4 8 2 0 2 6 2 8

W ith upper a n ti- coincidence 6 . 5 14. 2 24 10 . 3 4 0 5 1 2 5 5 only

W ith upper and low er a n ti- coincidence 1 . 0 6 . 5 3. 5 6 . 9 . 1 9 2 -, 59 2 . 1 ASSESSMENT OF Pu 239 LUNG BURDENS 145

Fig. 1 Large-area proportional counter, window dimension 15 x 30 cm г, for measuring low-energy X-rays Three flat counters are mounted directly together in common casing to give optimum background réduction by anti coincidence shielding (see Table 1)

at a distance of 10 cm from the centre of the counter window yields 1550 cpm. This figure corresponds to a counting efficiency of 26.5% for the incident X-rays, if one considers the geometry factor (12%) and the air absorption (43%, see [4]). The discrimination factor against gamma radiation of Co 60 is m ore than 1000.

3. CALIBRATION AND DETERMINATION OF THE OPTIMUM ARRANGEMENT OF COUNTERS FOR IN VIVO ASSESSMENTS

For calibration purposes two artificial lung lobes, made of plastic foam, were modelled on an actual lung. The plastic foam was then soaked with a solution of Pu239, the activity of which had been determined by alpha measurement. The purity of the solution had been checked by alpha and gamma spectroscopy. All told, the amount of solution was such that the average weight of a human lung (about 500 g for each lobe) had been reached. Afterwards, the plastic foam sponges were tightly sealed in plastic bags in order to avoid contamination (Fig. 2). Self-absorption in one of these models amounted to the factor of 3.9 for the X-rays of Pu239, as against a point- source sample of the same activity at a distance of 30 cm from the counter window. F rom a m ale corpse 89 kg in weight and 1.80 m in s ize the two lungs were removed through the abdominal cavity and replaced by the two models

10 146 R. EHRET et al.

Fig. 2

Plastic lung lobes before (left lobe) and after (right lobe) filling with plutonium solution Right lobe has space carved out for placing the heart

filled with plutonium. Then the abdomen was closed again. The counter described above was then used for determining the counting rates in various positions of the counter as well as at various distances from the surface of the body. Some of these positions with the corresponding results are shown in Fig. 3. In the arrangement of the counter above the centre of the chest (Fig. 3a) an increase in distance of from 2-5 cm of the counter from the skin reduced the counting rate by 19%. From this one concludes that thanks to the large area of the counter, the distance of the counter from the skin does not affect the measurement too critically. The final set-up, chosen on the basis of the evaluation of the measure­ ment, consisted of two proportional counting tubes arranged according to Fig. 3b above the chest to the right and left sides of the breastbone in a slight­ ly oblique position. (The vertical arrangement directly beneath the arms, Fig. 3c, increases the background slightly due to the attenuation of the anti­ coincidence set-up. Despite the decrease of efficiency, this arrangement may be preferred for monitoring female persons.) As the set-up on the back

10» ASSESSMENT OF Pu289 LUNG BURDENS 147

Fig* 3

(Explanatory notes please see next page) 148 R. EHRET e t al.

Fig- 3

Relative detection efficiency for various counter positions

(a) above middle o f chest (b ) sideways above chest (c ) on side of chest (d) above middle of back (e ) transversally above small o f the back

counter re la tiv e chest absorption counter position co u n t-ra te fa cto r + position facto r +"

(a ) 1 0 0 2 9. 5 2. 5

(b ) 123 17 3. 5

( c ) 98 15 5

(d ) 19 1 5 4 2. 5

(e ) 2 1 8 8 4

+ counts with lung lobes outside of chest counts with lung lobes inside of chest for same counter position

+ * counts with both lung lobes directly in front of counter counts with lung lobes outside of chest for respective counter positions reduces the counting rate by a factor of more than 10,due to the absorption in the shoulder blades, monitoring from the back is pointless.

4. LIMITS OF DETECTION

In the final arrangement of the two counters (Fig. 4) a total of 1 mc P u 239 in the two lung models yields 56 cpm at a background of 3 cpm in the steel- room. For a measuring time of 30 min with a confidence limit of 95% this would mean a detectable amount of Pu of 1.2X 10-8 c. Apart from the in vivo measurement of plutonium in the lungs the same set-up may of course be used also for the in vivo measurement of plu­ tonium ingested in the gastro-intestinal tract, proper calibration provided.

ACKNOWLEDGEMENTS

We should like to express our sincere thanks to Mr. E. Baumgartner, Electrical Engineer, for his reliable co-operation in setting up the equipment and making the measurements, and to Mr. R. Becker, Mechanical Engineer, for his speedy and careful construction of the component parts. ASSESSMENT OF Pu239 LUNG BURDENS 149

Fig. 4

Final arrangement of two proportional counters for in vivo measurements of plutonium lung burden Electronics consist of three transistorized high tension units, transistorized preamplifiers, pulse delay unit and com m ercial multichannel analyser For taking the picture, detectors and subject have been taken out of the steel-room to give better view of the whole apparatus

REFERENCES

Ш ROESCH, W.C. and PALMER, H. E. , “Detection of plutonium in vivo by whole body counting", Hlth Phys. 8 (1962 ) 773. [2] KIEFER, H. and MAUSHART, R. , "Nachweis von Pu*35" , Strahlenschutzmesstechnik (Abschnitt 3.629), Braun-Verlag, Karlsruhe (1964). . [3] RUNDO. J ., Discussion V. 3, Whole-Body Counting. IAEA. Vienna (1962) 295. [4] FESSLER, H ., KIEFER, H. and MAUSHART, R .. "Zur Messung von Quantenstrahlung im Energiebereich von 3-30 keV mit grossflâchigen Proportionalzâhlrohren;' Atompraxis 1 11 (1961) 401. [5] KIEFER, H. and MAUSHART, R. , "Determinationofplutonium-239bodyburdenusinggammaspectrometry with proportional counters", Whole-Body Counting, IAEA, Vienna (1962 ) 289.

[ 6 ] KIEFER, H. and MAUSHART, R .. "Large-area flow counters speed radiation measurements", Nucleonics 19 12 (1961) 51.

DISCUSSION

. (on the three foregoing papers.)

H. WYKER: I should like to ask Mr. Peabody to give us the energy range in which it is necessary to impregnate the bone of the skeleton used for a phantom with tissue-equivalent material. You stated that it was not neces­ sary at the low energies you used. I think this is because the attenuation in the bone overshadows the attenuation in the holes, whether filled with tissue-equivalent material or not. Impregnation may also be unnecessary at high energies, where the absorption in bone has decreased considerably. C.O. PEABODY: The attenuation produced by the non-impregnated bone at the energies we are using (up to 60 keV) is very large compared with that of the bone m arrow and the soft tissues of the chest. The addition of tissue- equivalent material irithe bone cavities would therefore produce no practical difference. At slightly higher energies, where the attenuation becomes less dependent on the effective atomic number, the difference in attenuation be­ tween the hard component.of the bone and the soft tissue becomes less. The effect of impregnation then becomes m ore important. We have not done any measurements on this but I would estimate that above 80 to 100 keV the presence of tissue-equivalent material in the bone cavities would be neces­ sary for the attenuation of the phantom bones to be sim ilar to that of live bones. There may be an upper energy limit at which impregnation again ceases to be important because of the small attenuation in bones and soft tissue, but I cannot guess what this might be. J. RUNDO (Chairman): Would any of the speakers like to comment on the other papers? There seems to be a difference of opinion between Dr. Maushart and-Mr. Peabody as to where the plutonium would be in the body. I think we could be fairly sure it would be between the two extremes of a single concentrated point in the chest and a general distribution through­ out the lung, but perhaps there are some people present who have information on the sort of distribution which is likely to be found in practice. Such in­ formation might be very valuable for those working in this field. C.O. PEABODY: My information has all been gleaned from medical colleagues, and it would appear that particles of plutonium o r plutonium oxide with diameters in the micron range will rapidly be distributed to all parts of the lung, some remaining, of course, in the upper bronchi. A ta la te r period, however, a considerable amount of the insoluble plutonium can be removed by the phagocytic process in the lung, via the lymph system, and will even­ tually appear in the lymph nodes around the junction of the trachea and the bronchi. I would think that some calibration should be done for this particu­ lar position as well as for the distributed, complete distribution throughout the lung, which Dr. Maushart described. ' I am wondering, in fact, whether Dr. Maushart has done any calibrations for point sources of plutonium in his model lung as well as for the widely distributed plutonium. R. MAUSHART: No, we have not used point sources, but with a real case of plutonium inhalation one does not know the distribution of the plutoni­ um. Even if one has calibrated for different distributions, one does not know which calibration to use. The best counter arrangement is, therefore, the one that gives the least variations for these different types of calibration.

151 152 DISCUSSION

J. RUNDO: Thank you, Dr. Maushart. This is perfectly true of course, but at the same time it would probably be desirable for the effect of the various distributions on the calibration factor to be examined, so that one can decide in practice what reliability one can attach to one's estimate of the content. • ■ W.N. SAXBY: I would like first to comment on the distribution of plutoni­ um in the lung. In most practical cases arising from inhalation of plutonium the individuals exposed are unlikely to be found and sent for whole-body assay for many days, by which time most of the very large particles will probably have been cleared upwards in the bronchial tree and swallowed and excreted. The remaining very fine particles will be moving towards a peripheral distri­ bution in the lungs as stated by Mr. Peabody. Secondly, Mr. Peabody suggested the use of Am241, as a built-in tracer, for assaying older, more highly irradiated plutonium. Of course, not all plutonium is in this category and care must be taken. 'Where Am241 is present, can Mr. Peabody be sure that the metabolism of plutonium is the same as that of americium? . . C.O. PEABODY: There is very little information available from which to estimate the extent to which plutonium and any americium which it contains w ill stay together in the human body after ingestion or inhalation. In our case, however, it is most probable that the plutonium would be inhaled in an in­ soluble particulate form such as РиОг. It is therefore most unlikely that the plutonium would become separated to any appreciable extent from its associated americium in the period between inhalation and measurement. BIOASSAY ssions A, 5 and

SAMPLING AND ANALYSIS FOR ASSESSMENT OF BODY BURDENS

J. H. HARLEY HEALTH AND SAFETY LABORATORY, UNITED STATES ATOMIC ENERGY COMMISSION, NEW YORK, N. Y., UNITED STATES OF AMERICA

Abstract — Résumé — Аннотация — Resumen

SAMPLING AND ANALYSIS FOR ASSESSMENT OF BODY BURDENS. A review of sampling criteria and techniques and of sample processing methods for indirect assessment of body burdens is presented. The text is limited to the more recent developments in the field of bioassay and to the nuclides which cannot be readily determined in .the body directly. A selected bibliography is included. The planning of a bioassay programme should emphasize the detection of high or unusual exposures and the concentrated study of these cases when detected. This procedure gives the maximum amount of data for the dosimetry of individuals at risk and also adds to our scientific background for an understanding of internal emitters. Only a minimum of effort should be spent on sampling individuals having had negligible exposure. The chemical separation procedures required for bioassay also fall into two categories. The first is the rapid method, possibly of low accuracy, used for detection. The second is the more accurate method required for study of the individual after detection of the exposure. Excretion, whether exponential or a power function, drops off rapidly. It is necessary to locate the exposure in time before any evaluation can be made, even before deciding if the exposure is significant. One approach is frequent sampling and analysis by a quick screening technique. More commonly, samples are collected at longer intervals and an arbitrary level of re-sampling is set to assist in the detection of real exposures. It is probable that too much bioassay effort has gone into measurements on individuals at low risk and not enough on those at higher risk. The development of bioassay procedures for overcoming this problem has begun, and this paper emphasizès this facet of sampling and sample processing.

PRÉLÈVEMENT ET ANALYSE POUR L’ÉVALUATION DE LA CHARGE CORPORELLE. L’auteur passe en revue les critères et techniques d'échantillonnage et les méthodes de préparation des échantillons appliquées pour dévaluation indirecte de la charge corporelle. Le mémoire porte exclusivement sur les plus récents progrès en matière d’analyse biologique, et sur les radioéléments qui ne peuvent pas être dosés directement dans l'organisme par une méthode simple. Il contient également un choix de références bibliographiques. Tout programme d'analyses biologiques devrait être conçu avant tout pour la détection des expositions élevées et inhabituelles et pour l’étude particulière des cas ainsi dépistés. Cette manière de procéder apporte le maximum de données pour la protection des individus exposés au risque et améliore les bases scientifiques dont on dispose pour l'étude des émetteurs internes. On ne devrait consacrer qu’un minimum de temps et de ressources au prélèvement d'échantillons sur les individus dont l’exposition est négligeable. Deux méthodes de séparation chimique sont employées pour l'analyse biologique: la première est la méthode rapide, dont le degré d'exactitude peut être faible et qui est utilisée pour la détection; la seconde est la méthode plus exacte qu'il faut employer après la détection de Texposition pour étudier l’individu exposé. L’excrétion,, qu'elle s'exprime par une fonction*exponentielle ou par une fonction de puissance, décroît rapidement avec le temps. Il faut donc connaître la- date de l’exposition pour faire une évaluation ou même seulement pour déterminer si l’exposition est significative. Pour cela, on peut prélever et analyser des échan­ tillons à des intervalles très rapprochés par une méthode de triage rapide. Mais le plus souvent, on rassemble des échantillons à des intervalles plus longs et un certain nombre de rééchantillonnages fixé arbitrairement facilite la détection des expositions réelles. Il est probable que, dans les programmes d'analyses biologiques, on a consacré jusqu'ici trop de temps et de ressources à des mesures sur les individus faiblement exposés au détriment de ceux qui encourent un risque plus considérable. On commence à mettre au point des méthodes d'analyses biologiques en vue de

155 156 J. H. HARLEY

résoudre ce problème, et le mémoire étudie précisément le prélèvement et la préparation des échantillons considérés sous cet angle.

ВЗЯТИЕ И ОБРАБОТКА ПРОБ. Представлен обзор критериев и методов взятия и об­ работки проб для непосредственной оценки содержания изотопов в организме человека. Об­ зор ограничивается рассмотрением последних достижений в области биологического анализа и радиоизотопов, которые не могут быть сразу определены в организме. К докладу прила­ гается выборочная библиография. При планировании программы по биоанализу следует обратить внимание на обнаружение высоких или необычных излучений и на тшательное изучение этих случаев при их обнаружении. Эта процедура дает максимальное количество данных для дозиметрии отдельных лиц, под­ вергающихся опасности, а также служит дополнительным научным обоснованием для понимания проблемы" внутренних источников излучения. Следует тратить минимум усилий-на отбор лиц с незначительным уровнем облучения. Процедуры по химическому разделению, необходимые для биоанализа, также делятся на две категории. Первая включает быстрые методы, возможно малой точности, применяемые для обнаружения излучения, вторая — более точные методы, необходимые для изучения ин­ дивидуума после обнаружения облучения Выделение — будь-то экспоненциальное или силовая функция — падает быстро. Необхо­ димо установить сроки облучения до какой-либо оценки, даже до решения того, является ли облучение значительным. Один метод состоит в частом взятии проб и анализе с быстрым отсеиванием материала. Чаще пробы берут через более длительные промежутки времени, и устанавливается средняя частота повторного взятия проб для обнаружения действительного об луч ен и я. Вероятно, что слишком значительные усилия при биоанализе тра-тятся на измерёния от­ дельных индивидуумов при небольшой опасности облучения и все же они не являются доста­ точными в отношении их при большей опасности.. Начата разработка процедур биоанализа для решения этой проблемы, и в докладе обращается внимание на этот аспект взятия проб и их о б р аб о т к и .

MUESTREO Y ANÁLISIS PARA EVALUACIÓN DE LA CARGA CORPORAL. En la memoria se reseñan los criterios y técnicas de muestreo y los métodos de tratamiento de muestras para evaluación indirecta de la carga corporal. El texto se circunscribe a las novedades más recientes en materia de análisis biológico, y a los núclidos que no pueden determinarse fácilmente de manera directa en el organismo. Se incluye una bibliografía selecta. ' Al prepararse un programa de análisis biológicos, debe atenderse principalmente a la detección de exposiciones elevadas o poco habituales y al estudio intensivo de tales casos, cuando se detecten. Este procedi­ miento proporciona el máximo de datos para.la dosimetría de personas expuestas a riesgos de irradiación y contribuye a'consolidar nuestros conocimientos científicos sobre los emisores internos. A las personas cuya exposición sea insignificante, sólo debe prestárseles un mínimo de atención. Los procedimientos de separación química necesarios para los análisis biológicos se dividen también en dos categorías. La primera comprende los métodos rápidos, tal vez de escasa precisión, utilizados para la detección^ La segunda abarca los métodos más exactos necesarios para el estudio del individuo después de detectarse la exposición. ■ . La ¿liminación por excreción, tanto según una función exponencial como de potencias, decrece rápida­ mente. Antes de proceder a una evaluación, incluso antes de decidir si la exposición es significativa, es preciso situar dicha exposición en el tiempo. Una manera de abordar este problema es efectuar muestras y análisis-frecuentes por un.método selectivo rápido. Pero es más corriente tomar las muestras a intervalos más largos y proceder a nuevos muéstreos discrecionales para ayudar a detectar las verdaderas exposiciones. Es probable que los trabajos de análisis biológicos dedicados a las personas poco expuestas hayan sido excesivos, e insuficientes los consagrados a las sometidas a exposiciones elevadas. Se han comenzado a per­ feccionar los procedimientos de análisis biológico para vencer esta dificultad y en la memoria se destaca este aspecto de la toma de muestras y del tratamiento de las mismas. • •

My topic of sampling and analysis includes two-thirds of our overall problem, the third part being evaluation of the data. I hope I can stay far enough away from evaluation to avoid conflict with other speakers, but the ASSESSMENT OF BODY BURDENS 157

selection of procedures'for sampling and analysis depends on the use that is to be made of the results. I will also try to avoid discussing the use of sampling and analysis of biological material as an adjunct to air sampling and confine myself to the assessment of body burden. The broad field of assessment of radioactive body burdens in man is referred to increasingly as bioassay. Following the example of the report by LISTER [1], I will use the term in this paper. It might be of interest to note that for about ten years, there has been an informal group in the United States that meets once a year to discuss mutual problems and to pre­ sent new techniques in bioassay. The proceedings of these meetings are listed in the bibliography at the end of this paper. The primary purpose of a bioassay programme is the detection and the evaluation of significant exposures. Most high accidental exposures are recognized from the physical facts of the case. The occasional high acci­ dental exposure that is discovered by a bioassay and the high chronic ex­ posure are the real problems in a sampling and analysis programme. Once a high exposure has been found, the utmost effort should be expended to document the case as completely as possible and to obtain both an evaluation of the exposure and scientific data on the behaviour of the particular element. The body burden of many gamma emitters may be measured with a whole-body counter, or even with a simpler device, usually at a small frac­ tion of the accepted permissible level. With alpha and beta emitters, how­ ever, an attempt must be made to assess the burden indirectly. This means measuring the retention of the nuclide by sampling and analysis of excreta. Since retention and excretion are direct opposites, we must use empirical relationships for assessment. If we assume that we have analysed a urine sample and have obtained a measurement for a specific nuclide, we are still not in a position to decide what the corresponding body burden is. This conversion is highly dependent on the time since exposure, the route of intake and the chemical and physical forms of the material*-. This means, for example, that the evaluation of a high accidental exposure really requires a considerable amount of back­ ground information on the conditions surrounding the accident. Even when these data are available the usual metabolic differences from individual to individual may obscure the results.

Standards .

Once the body burden has been assessed it is usually compared with the maximum permissible body burden specified by the International Com­ mission on Radiological Protection (ICRP). From time to time questions have arisen as to the suitability of this comparison, based on the following points:

* For example, JACKSON [2] has published estimates of urinary uranium excretion at various times for a body burden of one-tenth the maximum permissible. Immediate sampling would give 4200 Mg/1, one week after exposure 170 Mg/1 and one month after exposure 26 Mg/1, when exposure was to the in­ halation of soluble natural uranium compounds. The same time periods would show 22, 21 and 17 Mg/1 when exposure was to the inhalation of in - soluble natural uranium. . ' 158 J. H. HARLEY

1. The basic human or animal excretion and retention data are for single acute exposures. 2. The single acute exposures are converted to chronic exposures using an exponential excretion model. 3. The fraction of the body burden which exists in the critical organ is most frequently determined from concentrations of the stable elements at equilibrium following chronic exposures. ' The overall comparison between the calculated bod}' burden and the ICRP value seems valid for chronic exposure, since any error is probably less than that involved in the body-burden assessment. Acute exposures will tend to appear very alarming when compared with ICRP values.' The only alternative is to estimate the individual organ burdens and calculate the ra­ diation dose for each case. The accuracy of this procedure is also likely to be low.

Regulations ■ •

To introduce another complication, it might be said that there are now two general reasons for carrying on a bioassay programme. One is an at­ tempt to prevent somatic damage to the individual worker and the other is to assure compliance with existing regulations. The approach to the first requirement can be empirical and flexible, although difficult, since each case must be evaluated by an experienced scientist. The second require­ ment can-be based only on numerical standards; judgement, is not allowed and each exposure must be evaluated on the basis of the most conservative conditions possible. This approach does not require highly experienced personnel but does lead to very restrictive, levels of control. In general, regulatory groups have specified maximum permissible occupational body burdens for various nuclides which are numerically equal to the recommendations of the ICRP. The means of assessing the body burden are not described. In the United States, for example, bioassay is recommended for assessing body burden but no procedures or recommended levels for excreta are specified. It is for the operator to show that the body burdens in his factory are below the permissible limits. To do this effec­ tively, he requires guidance. . ■ Now that we have established that we face an impossible task, let us see what possible steps we can take to approximate the desired information. Most of the burden falls on the process of sampling.

SAM PLING -

The most difficult aspect of bioassay sampling is planning a programme which will provide adequate screening to detect accidental exposures or sig­ nificant chronic exposures. Once it is known .that, an exposure exists, there is no problem. All excreta should be collected and all samples should be analysed, either individually or as pooled specimens, until the results indi­ cate that spaced samples are sufficient. ASSESSMENT OF. BODY BURDENS 159

Most installations consider urinary sampling as the body screening procedure. Partially this is based on tradition and partially it is based on the concept that faecal material has never really been in the body. Urinary excretion may be considered to follow an exponential curve, a series of exponentials, or a power function. The series of exponentials seems most probable, since this may be interpreted as a- series of body compartments or organs having.different excretion rates. When there are several exponentials in the series the curve can be approximated very well by a power function. Regardless of the form of the curve, the total dose at any time may be estimated by integrating the curve from the time of exposure. A sipiilar procedure may be carried out for individual compartments by breaking down the series of exponentials into individual components. ( The initial body burden from an acute exposure computed by extrapo­ lating a retention curve back to the time of exposure is a useful quantity but is not directly comparable with the IC R P maximum perm issible body burden. As mentioned before, the ICRP value is based on chronic exposure and is not applicable. For this reason there is a tendency for high accidental ex­ posures to be studied individually and the dose computed for the particular case. Obviously, this amount of effort is not justified in the case of minimal exposures. There has been some work on the application of com­ puters to the problem , but the process should, s till be lim ited to estim ating the dose to workers with significant exposure. The exceptions to all of these problems in estimating body burden are the èlements that equilibrate rapidly in the body and do not concentrate in particular organs. Tritium and Csi37 can be readily assessed,as they appear to be excreted, from a single compartment as a single exponential. Thus, their concentration in urine will approximate to their concentration in body fluids quite closely. It is frequently not realized that the curve which would represent the penetration of a particular element into the body is a rough m irror image of an excretion curve.. Thus, excretion from a particular organ with a long biological half-life usually indicates that comparable times are required to' get material into that organ foilowirig an intake by inhalation or ingestion. This particular point is important in sampling following an accidental ex­ posure. The material being excreted at high concentrations has never pene­ trated to the less accessible body organs. The excretion may be merely clearance of the upper respiratory tract or the blood stream. The actual dose rate being delivered may be high but the dose may not be critical since the time is short or the organ receiving the dose may not be radiosensitive. On the other hand, excretion with a longer biological half-life indicates excretion from organs further along a metabolic path. A lower rate of ex­ cretion (longer retention) if ^combined with greater radiosensitivity of the organs concerned may produce the most critical dose even though the nuclide concentrations are much lower. Thus, for purposes' of dose evaluation, we may have only a slight interest in the early period of excretion even though thé levels are high. Such data are more helpful in estimating air concentrations than dose. It should be emphasized, of course, that information not required for dose 160 J. H. HARLEY assessment may still be very valuable in increasing our knowledge of the metabolism of a particular element.

The sampling programme

The planning of a bioassay screening programme must, of course, be based on the technical requirements, but the cost in money and in manpower must also be considered. The overall programme must have a favourable balance of value to cost. The decisions required are the type of sample, the frequency of sampling and the personnel to be sampled. Since this is a screening programme, the sampling should not be so burdensome that co-operation is difficult. ' The selection of sample type is still a problem. While technical con­ siderations are paramount, it. cannot be ignored that a urine sample is easier to obtain than faeces, and that a spot urine sample is easier to obtain than a 24-h output. There is also art analytical advantage,, first, in the smaller urine sample and, second, in urine samples as compared to faeces. These advantages, of course, are based on the assumption that chemical processing is required rather than merely measurement of the gamma spectrum. . The maximum permissible body burden of any gamma-emitting nuclide is readily measured in a whole-body counter or even in a simpler unit in­ volving less shielding or smaller crystals. The saméis not true when measurements are made of excreta. In my opinion, faecal sampling as a means of estimating body burden has been greatly over-emphasized. The general route of intake for material that appears in the faeces is ingestion, the time in the body is certainly short, and the gut is not very radio-sensitive. The use of whole-body counting for- measuring faeces samples therefore seems to destroy many of the advantages of the counter. The desired value is the retention of the material, not the excretion. If they can both be measured readily it seems best to measure the amount retained. It must be remem­ bered that the whole-body counter measures the total-body burden and not the amount in a specific organ. I believe we are reduced to urinalysis as a screening procedure for detecting accidental or high chronic exposures to alpha and beta emitters. We have seen the difficulties that exist in evaluating the data, but the al­ ternatives are less attractive. . To describe an exposure level from a bioassay result we need two things, the general shape of the excretion curve and the time since exposure. These cannot be known exactly, but they can probably be approximated closely enough. . The shape of the excretion curve is obtained empirically from measure­ ments on high exposure cases. The route of exposure and chemical and physical form of the element must be the same as for the case being con­ sidered. This means-that a library of exposure data is required. Some of it exists already, but no opportunity should be lost to add to our knowledge. The time since exposure is not known; it must therefore be approxi­ mated with reasonable accuracy. The best approach to this is by frequent sampling. Since all personnel cannot be sampled continually, a degree of risk is required. Workers with the highest potential exposure should be ASSESSMENT OF BODY BURDENS 161

sampled at the beginning of every work week, and other workers less fre­ quently in proportion to the potential exposure. The weekly samples are a starting point, and frequency may be decreased as experience dictates. If the analytical load is too great, the lower exposures.-should be cut first. With samples taken at the start of the work week, the very short half­ time components will have disappeared. The half-period of the shortest remaining component must be of. the order-of many hours. The exposure, if any, must have, taken place during the preceding week, so that the time scale is relatively fixed. , The sampling programme described will cover both the undetected ac­ cident and the high chronic exposure. The. former will appear as a sharp rise in excretion, the latter as a slow rise. The programme will not be a substitute for air monitoring and it is not intended for any purpose other than the assessment of body burden. ■ Another approach to bioassay sampling is the measurement of frequent samples from all exposed workers. An arbitrary re-sampling level is set, above which a new sample is collected and analysed for the particular workér. Like all other samples, the one that is, the criterion for re-sampling is dif­ ficult to interpret unless the time since exposure is known. If the re­ sampling point is chosen to detect almost all of the exposures that are of significance, this choic.e will require1 re-sampling of a large number of in­ dividuals without exposure. It is probable that less effort is required to. carry out the weekly sampling previously described. In setting up a sampling programme considerable attention should be paid to maintenance personnel who enter contaminated areas. They are very likely to be.in the work areas at times when equipment is not operating properly and thus when relatively high exposures are possible. It has been our experience that maintenance workers frequently have high,' short-term exposures and that they should be included in the most frequently-sampled group. . .

Mechanics of sampling .

Once the sampling programme has been decided upon, the actual m e­ chanics of collection, storage and treatment of samples seem relatively simple. It is necessary to collect the samples in such a way that the radio­ nuclide does not become segregated and so that the samples in themselves do not become objectionable. ' . The prevention of sample contamination is probably best attained by sampling away from the work area, either at home or in the plant dispensary. It is necessary to give plant personnel general instructions on washing and other aspects of sam pling and not to take their understanding fo r grante.d., A few references on sampling techniques are given in the bibliography and, of course, the original recommendations in the joint WHO-FAO Report on Radiochemical Methods of Analysis [3] are still valid. .

ANALYSIS ■ • ; •

Analytical methods required for bioassay screening must bè rapid and simple and, if possible, economical. Since any of the bioassay programmes 162 J. H. HARLEY described require frequent sampling, the number of samples is large and results must be available quite rapidly. Fortunately, the requirement for accuracy is very small. Ordinarily the results, should be known to about a factor of two. Such data are entirely sufficient for sorting out those individuals with possible exposure from those who are of no interest. A typical method that might be considered adequate for speed and simplicity is the determination of natural uranium in urine by the fluorimetric technique. Such methods do not exist, however, for the majority of other radionuclides. Almost all industrial exposures are to single nuclides or, at worst, single m aterials, such as natural uranium. Thus, the bioassay requirement for radiochemical purification is slight. The analytical problem most often is only the isolation of the radioactivity from the large amounts of inert bulk m aterial of the sample matrix. Thus, the analyst engaged in routine bioassay is in a very fortunate po­ sition, since the methods have very minimal needs for accuracy, separation efficiency, and frequently also for sensitivity. All that is needed is a screening process to pick out the individuals with significant exposure. The same laboratory that is involved in the routine monitoring of radio­ activity in man will ordinarily be required to evaluate the significant ex­ posures and may also be carryin g out experim ental studies on uptake, r e ­ tention or excretion of the same nuclide. Thus, it is not unusual for a la­ boratory to require two analytical procedures, one procedure of high ac­ curacy regardless of time and effort and the second of low accuracy but simple and rapid enough for routine use on a large number of samples. These two requirements are not in conflict, but it is a rare case when the more exact method is rapid enough for routine operation. For moni­ toring purposes it is frequently considered that a sensitivity of 2 5% of the established level is satisfactory and that errors of -50 to +100% are ac­ ceptable. Such a technique, however, would certainly not be satisfactory when trying to evaluate an excretion pattern following exposure. . The attached bibliography lists a number of bioassay methods which have appeared since thé joint WHO-FAO Report issued in 1959. The listing is not exhaustive, but the literature cited is considered to be reliable and adds to our background of information on analytical methods for bioassay. No attempt is made in this discussion to describe individual techniques, since' progress has been a matter of modest improvements rather than revolutionary changes. . From time to time there is pressure at a national or international level for the standardization of analytical methods. Such standardization would be of little value and would certainly tend to halt useful analytical develop­ ment. There is no need for all laboratories to be using the same technique. It is far better that they should be' able to obtain the same results when ana­ lysing the same samples by different procedures. The problem of the intercomparison of standards and samples between laboratories was the subject of an IAEA meeting at the end of 1963. More limited meetings have been taking place for several years in an attempt to determine if data obtained at different laboratories are directly intercom­ parable. The discussions and the resulting exchange of samples have sti- ASSESSMENT OF. BODY BURDENS 163 mulated analytical progress in this field and in many cases have improved the quality of results. . •• •• ' ■ ■ . In work on the development of methods it is necessary to have at hand samples with known concentrations of the nuclide being sought. Since the chemical form of the nuclide is frequently very important, the best samples are ones where the nuclide has been metabolized and the sample is essentially a "natural" material. Such samples have been readily available for fallout programmes but are relatively difficult to obtain in studies of radioactive contamination of man from industrial sources. I would like to emphasize that the present methods of'sampling and ana­ lysis for assessing radioactive body burden in man have been developed to fill a need. In the past, this need has been based on indications of high ex­ posures of a significant number of workers to particular radionuclides. Thus, when we turn to the literature we find numerous methods for measuring uranium, thorium, tritium, strontium-90, plutonium and caesium-137. We do not, in general, find comparable methods for many hundreds of other nuclides. This lack of 'methods for these other nuclides has not been critical, since the number of workers exposed and the degree of exposure have been minimal. At the present time, however, there is a tendency for those writing regulations to specify permissible exposures to a broad spectrum of radionuclides. Chemists who have been involved in the analysis of nuc­ lides of radiological significance are receiving an increasing number of re­ quests for information on less common radionuclides. ■ _ In the United States at the present time there is an attempt by the Atomic Energy Commission to fill this need. The Commission is giving active en­ couragement to programmes of evaluating the present methods for the com­ mon nuclides and of developing techniques for the less common ones. Un­ fortunately, in the latter case the immediate value of the methods is dubious, since there are few data on the biological behaviour of these nuclides. .

SUMMARY ■ ./ ■ ' ' '

The analysis of bioassay samples is seldom a major problem compared to those involved in sampling and in interpreting bioassay data. The major requirement is for simple, rapid methods that can handle large numbers of routine samples, A sampling programme should be designed to detect unexpected or ac­ cidental exposures and to estimate the level of chronic exposures. When a significant exposure is found, the case should be followed in complete detail to evaluate the dose to the individual. .

’ " . REFERENCES . •

[ 1 ] LISTER, B. A. J. , Problems and methods of bioassay programmes,UKAEA Report AERE-R 4202 (1962). [2] JACKSON, S., The interpretation of data on radionuclides in urine, Part 1: Uranium.UKAEA Report AHSB (RP)R15, [3] WORLD HEALTH ORGANIZATION, Methods of radiochemical analysis, Technical Report Series Number 173 (1959) (Also available as FAO Atomic Energy Series Number 1). • 164 J. H. HARLEY

BIBLIOGRAPHY

DATA INTERPRETATION

VENNART, J. , Assessment of the internal hazard, Ann. Occupational Hyg. 4 (1961) 127-33.' LIPPMAN, M. et al. , Significance of urine uranium excretion data, USAEC Report HASL-120 (1960). FISCHOFF, R. L. , Correlation between two-stage air sampling data and the excretion of uranium in urine, USAEC Report NLCO-869 (1963).

COLLECTED METHODS "

BROOKS, R. O. R. , Collected laboratory procedures for the determination of radioelements in urine, UKAEA Report AERE-AM-60 (1960). UNITED STATES ATOMIC ENERGY COMMISSION, Health and Safety Laboratory. Manual of standard procedures USAEC Report NYO-4700 (1957). . HURSH, John B. , Chemical methods for routine bioassay, USAEC Report AECU-4024 (1958). JEANMAIRE, L. and JAMMET, H. , Radiotoxicologic analysis of urine, Ann. Radiol. 2 (1959) 703-22. MILLIGAN, M. F. , CAMPBELL, E. E. , EUTSLER, B .C ., McCLELLAND, Jean and MOSS, W. D., Analytical procedures of the industrial hygiene group, USAEC Report L'A-1858 (2nd ed. ) (1958). MILLIGAN, M. F. , Chemical methods in the evaluation of human contamination, Symposium on Technical

Methods in Health Physics, RísíZÍ, Denmark (May 25-28, 1959) 102-9, USAEC Report TID-7577. SCHULTE, К. E. and HENKE, G. , The determination of radioisotopes in urine, Deutsche Apotheker Ztg. 100 (1960) 700-7; also UKAEA Report AERE-Trans-869 (1960). SHIPMAN, W. H. and WEISS, H. V. , Analytical procedures at the US Naval Radiological Defense Laboratory for the determination of certain radioelements in urine, USAEC Report UNNRDL-TR-451 (1960).

UHLMANN, W. , The determination of radioactive isotopes in urine (Bibliography), USAEC Report VDIT - 6 6 (1 9 6 2 ). WOODMAN, F. J. , Critical survey of methods for the determination of radionuclides in urine. Symposium on Health Physics in Nuclear Installations, Ris^, Denmark (May 1959) 105-14, USAEC Report TID-7577. WORLD HEALTH ORGANIZATION, Methods of radiochemical analysis, Technical Reports Series No. 173 (1959); (Also available as FAO Atomic Energy Series No. 1)/ PROCEEDINGS OF THE BIOASSAY AND ANALYTICAL CHEMISTRY MEETING

HOOVER, Reynold L. , Proceedings of Bio-assay and analytical chemistry meeting (Oct. 6 and 7, 1955) USAEC Report NLCO-595. LOS ALAMOS SCIENTIFIC LABORATORY, Proceedings of second annual meeting on bioassay and analytical chemistry (Oct, ll and 12, 1956), USAEC Report WASH-736 (1957). No proceedings of the third meeting were issued. UNIVERSITY OF ROCHESTER, Proceedings of the fourth annual meeting on bio-assay and analytical chemistry (Nov. 3 and 4, 1958) USAEC Report WASH-1023 (1959). OAK RIDGE NATIONAL LABORATORY, Proceedings of the fifth annual meeting on bio-assay and analytical

chemistry (Oct. 1 and 2, 1959)/ USAEC Report TID-7591. - SANDIA CORPORATION, Albuquerque, New Mexico, Sixth annual meeting on bio-assay and analytical chemistry (Oct. 13-14, 1960), USAEC Report TID-7616 (1962).

TRITIUM

U. K. ATOMIC ENERGY AUTHORITY, Industrial Group, Analytical methods for the determination of Tritium in urine (2¿quid scintillation method), UKAEA Report PG-162 (1960). GIBSON, J. A. B. , Liquid scintillation counting of tritium in urine, Phys. Med. Biol. 6^(1961) 55-64. SANDALLS, J. , A method for routine determinations of tritium in urine using a coincidence liquid scintillation counter, UKAEA Report AERE-R-3716 (19’61). PEETS, Edwin A. , FLORINI, James R. and BUYSKE.Donald A. , Tritium radioactivity determination of bio­ logical materials by a rapid dry combustion technique, Anal. Chem. 32 (1960) 1465-8. MYERS, L. S. and ROSENBLUM, Cynthia, A rapid method for determining tritium water in urine following acute exposure, USAEC Report UCLA-504 (1962) 14. ASSESSMENT OF BODY BURDENS 165

URANIUM .

JEANMAIRE, L. and JAMMET, H. , Determination of uranium in urine, Commissariat à l’Energie Atomique Report 1226 (1959). WHITSON, Т. С. and KWASNOSKI, T. , The determination of alpha activity associated with sub-microgram quantities of enriched uranium in urine, J. Am. Ind. Hyg. Assoc. 20 (1959) 169-74. U. K. ATOMIC ENERGY AUTHORITY, Production Group, Analytical methods for the determination of uranium

in urine, UKAEA Report PGR-6 8 (W) (1959). _ , ROYSTER, George W., Jr. , Electrodeposition of uranium from urine, Health Phys. 2 (1960) 291-4. RONTEIX, Claude and HUGOT, Gilbert, Dosage fluorimetrique de l'uranium urinaire, Commissariat â l’Energie Atomique Report 1706 (1960). DIETRICH, William C. , CAYLOR, John D. and JOHNSON, Eric E. , Separation of uranium from urine by a tri-n-octylphosphine oxide column and an automation of the procedure, USAEC Report Y-1322 (1960). WELFORD, George A. , MORSE, Robert S. and ALERCIO, John S. , Urinary uranium levels in non-exposed individuals, J. Am. Ind. Hyg. Assoc. 21 (1960) 68-70. U. K. ATOMIC ENERGY AUTHORITY, Production Group, Analytical method for the fluorimetric determination of uranium in biological material, UKAÉA Report PG-143 (1960) 12p. WHITSON, T. C. and KWASNOSKI, T. , A semi-automatic apparatus and method for alpha activity determi­ nation in urine, USAEC Report TID-16090, 14 p.

MISCELLANEOUS

MARRIOTT, J. E. , The determination of iodine-131 in urine, Analyst 84 (1959) 33-7. ARKELL, G. M. and MORGAN, A. , The determination of cesium-137 in urine, UKAEA Report AERE-R-3675 (1 9 6 1 ). U. K. ATOMIC ENERGY AUTHORITY, Industrial Group, The determination of 210 polonium in urine and other biological materials, UKAEA Report IGO-AM/W-167 (1958). FOURNIGUET, H. , JEANMAIRE, L. and JAMMET, H. , Dosage du radium dans l’urine, Commissariat à l’Energie Atomique Report 1228 (1959). ¡’ DUPZYK, Isabelle and BIGGS, Max W. , Urinalysis for curium by electrodeposition, USAEC Report UCRL-6168 (1 9 6 0 ). BRUNO, Gerald A. and CHRISTIAN, John E. , Determination of carbon-14 in aqueous bicarbonate solutions by liquid scintillation coimting techniques. Application to biological fluids. Anal. Chem. 33^ (1961) 1216-18. BOKOWSKI, D. L. , The radiochemical determination of americium in the presence of plutonium in urine, USAEC Report RFP-279 (1961) 18. PERKINS, R. W. , A tentative bioassay procedure for the measurement of Np237, USAEC Report HW-67067, (1960) 3 p. BLACK, Stuart C. , Low level polonium and radiolead analysis, Health Phys. 1. 1 anc* 2 (1961) 87-91. HÜRSH, John B. and LOVAAS, Arvin, A device for measurement of thoron in the breath, USAEC Report UR-619 (1962) 28 p. , SAMSAHI, K. , Some chemical group separations of radioactive trace elements, USAEC Report AE-82 (1962). ARGONNE NATIONAL LABORATORY, Proceedings of the seventh annual meeting on bio'assay and analytical chemistry (Oct, 12 and 13, 1961), USAEC Report ANL-6637. E. I. DU PONT DE NEMOURS & COMPANY, Aiken, South Carolina, Papers presented at the eighth annual meeting on bio-assay and analytical chemistry (Oct. 18 and 19, 1962), USAEC Report DP-831. GENERAL ATOMIC, Division of General Dynamics, San Diego, California, Ninth annual meeting on bio-assay and analytical chemistry (Oct. 10 and 11, 1963) (Procedures not available).

SAMPLING AND DIRECT MEASUREMENT

GOMM, P. J. and TRUMP, M. R. , A comprehensive punch-card system for the routine biological sampling of personnel, UKAEA Report AERE-R-3448 (1960). . HAYES, R. L. , A method for collection and radioassay of specimens of human stools. Am. J. Clin, Pathol. 39 (1963) 203-5. STONE, G. F. , A device for'the collection of fecal specimens, USAEC Report CF“59~11“8, (1959) 7 p. BUCHANAN, Donald L. and SAMPSON, Lowell T. , Incineration of fecal specimens for radioactivity measure­ ment, J. Lab. Clin. Med. 59 (1962) 169-73. 166 J. H. HARLEY

CLAPHAM, W. F. and HAYTER, C. J. , The measurement of gamma emirting isotopes in feces, Phys. Med. Biol. 7 (1962) 313-17. PERKINS, R. W. , A large well crystal for the direct measurement of trace amounts of radioisotopes in en­ vironmental samples, Health Phys. 1_ 1 and 2 (1961) 81-6.

HILL, C. R. , An a-spectrometer for normal biological levels of activity, Health Phys. 8 (1962) 91*2. * WEST, С. M. and REAVES, J. P. , Use of statistics in an applied health physics program, USAEC Report Y-KB-7, (1963) 18 p. ‘ Г

PLUTONIUM , ■

WEISS, Hërbert V. and SHIPMAN, William H. , Radiochemical determination of plutonium in urine, Anal. Chem. 33 (1961) 37-9. -. ' . FOREMAN, H. , MOSS, W. and LANGHAM, W. , Plutonium accumulation from long-term occupational ex­ posure, Health Phys. 2 (1960) 326-33.

SANDERS, S., MARSHALL and LEIDT, Sarah C. , A new procedure for plutonium urinalysis. Health Phys. 6_ (1961) 189-97. . . ' DALTON, J. C. , The determination of plutonium in urine by direct phosphate precipitation and autoradiography, UKAEA Report PG-284, (1962) 30 p. . ■ BRUENGER, F. W. , STOVER, B. J. and ATHERTON, D. R. , Solvent extraction of plutonium with primary amines. Some applications to plutonium extraction from biological material, especially urine, USAEC Report COO-226 (1963) 58-65.

STRONTIUM ' , ,

ZBORIL, V. , SEBESTIAN, I. , TRNOVEC, T. and DURCEK, K. , The determination of radiostrontium in urine, USAEC Report AEC-tr-4222 (1960). U.K. ATOMIC ENERGY AUTHORITY, Production Group, Analytical method for the determination of strontium-89 and strontium-90 in urine, UKAEA Report PGR-99 (W) (1960). . JEANMAIRE, Lucien and MICHON, Georges, Methods for the determination of radioactive strontium in bio­ logical substances, Bull. Soc. Chim. France 2 (1963) 413-15. .

THORIUM ■

JEANMAIRE, L. and JAMMET, H. , Determination of natural thorium in urine, Commissariat à l’Energie Atomique Report 1227 (1959). HELMAN, A. H. , The determination of thorium in urine, UKAEA Report PG-178 (1961). • SILL, C. W. and WILLIS, C. P. , Fluorometric determination of submicrogram quantities of thorium, Anal. Chem. 34 (1962) 925. ’

' DISCUSSION

N. I. SAX: Has any thought been given to the use of the bioassay tech­ niques described in this paper and the papers in these proceedings by S. Jackson and N. Taylor, "A survey of the methods used in the UKAEA for the determination of radionuclides in urine", and P.R. Kamath et al."Recént radiochemical procedures for bioassay studies at Trombay" in connection with public health? I am thinking specifically of the movement of nuclides and contamination from reactors, fuel reprocessing plants, factories and so on to surrounding populations. J. H. H ARLEY: I do not know of any application of bioassay to the popu­ lation. . The protection of the public depends on restrictions on waste disposal. G. BUTLER (Chairman): In my experience it is not possible to carry out bioassay to any great extent on the public. Usually the public is protected ASSESSMENT OF BODY BURDENS 167 by environmental monitoring following upon an adequate design of the plant and its disposal facilities. Would anyone elsje .like, to contribute? P.R. KAMATH: As far as my country is concerned, we have not arrived at the stage where it is possible or necessary to make bioassay measure­ ments on the general population, but we are planning something of this kind in areas with a high natural radiation background. . H. WELLMAN: I should like to comment on a rather different point. Many elements are excreted endogenously. For many years Ca excretion in faeces was thought not to be important, but it is now known to be appreciable. Likewise, there is data indicating that endogenous excretion of Sr is signi­ ficant. Furthermore, indirect data on infants have given rise to speculation that endogenous excretion may be very important. Cs and К may also be excreted by the faecal route. With such sparse data on most non-absorbable nuclides it is probable that eventually endogenous excretion will play an im ­ portant role in exposure assessment. ‘ J.H. HARLEY: I realize that endogenous excretion may be important. Faecal excretion, however, does not distinguish between endogenous and exogenous, while urinary excretion always indicates metabolized material. I might add that bioassay can only provide an index to the body content, but there is no indication that there is any virtue in measuring endogenous faecal excretion per se.

A SURVEY OF THE METHODS USED IN THE UNITED KINGDOM ATOMIC ENERGY AUTHORITY ' FOR THE DETERMINATION OF RADIONUCLIDES IN URINE

; S. JACKSON AUTHORITY HEALTH AND SAFETY BRANCH, RADIOLOGICAL PROTECTION DIVISION, U. К. A. E. A, HARWELL, ENGLAND ; AND . ~ . . N. A . TAYLOR ' , HEALTH PHYSICS BRANCH, U. K.A.E.A, ALDERMASTON, ENGLAND.

- Abstract — Résumé — Аннотация — Resumen

A SURVEY OF THE METHODS USED IN THE UNITED KINGDOM ATOMIC ENERGY AUTHORITY FOR THE DETERMINATION OF RADIONUCLIDES IN URINE. With the co-operation of analytical, health physics and medical staff, a survey has been made throughout the United Kingdom Atomic Energy Authority of current practice in urine analysis for radionuclides. • • The elements which are of greatest importance in the UKAEA programmes of urine analvsis are plutonium, uranium, tritium and fission products (notably strontium-90 and caesium -137). Other radionuclides dealt with are polonium-210, radium-226, protactinium-231, phosphorus-32, carbon-14, sulphur-35, americium-241 and thorium. , In assessing the function and scope of urine analysis, the factors affecting the choice of a suitable analyti­ cal method are discussed. The sensitivity and precision of the method must be adequate in relation to the maximum permissible body burden, the excretion rate and the sampling frequency; the method must be suffi­ ciently specific or selective to eliminate the possibility of interference by other radionuclides; and finally, the cost must be assessed in relation to all these factors, and also to the speed and convenience of the method. The sensitivity required for each radionuclide is calculated from the maximum permissible body burden by applying a representative urinary excretion rate, based on the best data available. The methods of urine analysis which are currently used in the United Kingdom Atomic Energy Authority are fully described. • According to the calculated requirements for sensitivity, the best methods in use are capable of quanti­ tatively detecting, from urinary excretion, internal contamination with one-tenth of the maximum permissible body burden of any of the above radionuclides, and with one-hundredth of the maximum permissible body burden in the case of enriched uranium, tritium, polonium-210 or caesium -137.

ENQUÊTES SUR LES MÉTHODES EMPLOYÉES DANS LES ÉTABLISSEMENTS DE L'AUTORITÉ DE L'ÉNERGIE ATOMIQUE DU ROYAUME-UNI POUR DOSER LES RADIONUCLÉIDES DANS LES URINES. Avec la collaboration des analystes,des radioprotectionnistes et des médecins faisant partie du personnel, on a fait une enquête dans tous les établissements de l’Autorité de l'énergie atomique (AEA) du Royaume-Uni sur les méthodes normale­ ment suivies pour doser les radionucléides dans les urines. > Dans les programmes d’analyses des urines exécutés par l’AEA, les éléments les plus importants sont le plutonium, l’uranium, le tritium et les produits de fission (en particulier le strontium 90 et le césium 137). Au nombre des autres.radionucléides auxquels on s’intéresse figurent le polonium 210, le radium 226, le protactinium 231, le phosphore 32, le carbone 14, le soufré 35, Г américium 241 et le thorium. Lorsqu'ils définissent le rôle et l'importance de l'analyse des urines, les auteurs étudient les facteurs qui influent sur le choix d'une méthode analytique appropriée. La sensibilité et la précision de la méthode doivent être bonnes, eu égard à la charge corporelle maximum admissible, à la vitesse d'excrétion et à la fréquence des prélèvements d'échantillons; la méthode doit être suffisamment sélective pour éliminer toute possibilité d’interférence due à d’autres radionucléides; enfin, le prix de revient doit être déterminé, eu égard à tous ces facteurs et également à la rapidité et à la commodité de la méthode.

169 170 S. JACKSON and N. A. TAYLOR

A partir de la charge corporelle maximum admissible, les auteurs déterminent la sensibilité voulue pour chaque radionucléide, en appliquant un taux d'élimination urinaire représentatif, calculé à l’aide des meilleures données disponibles. Les méthodes d’analyse des urines que l'on applique à présent dans les établissements de ГАЕА font l'objet d’une description complète. D’après les critères de sensibilité ainsi déterminés, les meilleures méthodes actuellement employées permettent de déceler quantitativement, à partir de l’élimination urinaire, une contamination interne égale à un dixième de la charge corporelle maximum admissible de n’importe quel radionucléide mentionné ci- dessus, et une contamination interne égale à un centième de la charge corporelle maximum admissible lorsqu'il s’agit d’uranium enrichi, de tritium, de polonium 210 ou de césium 137.

МЕТОДЫ, ИСПОЛЬЗУЕМЫЕ В УПРАВЛЕНИИ ПО АТОМНОЙ ЭНЕРГИИ СОЕДИНЕН­ НОГО КОРОЛЕВСТВА ДЛЯ ОБНАРУЖЕНИЯ РАДИОИЗОТОПОВ В МОЧЕ. В сотрудничестве со специалистами по аналитической химии, радиационной безопасности и медицинским персо­ налом были рассмотрены практические методы анализа мочи на радиоизотопы, используемые в Управлении по атомной энергии Соединенного Королевства. ’ К элементам, имеющим наибольшее значение по программе Управления, относится плу­ тоний, уран, тритий и прдукты деления (в особенности стронций-90 и цезий-137). Кроме того, рассмотрены такие радиоизотопы,как полоний-210, радий-226, протактиний-231, фосфор-32, углерод-14, сера-35, америций-241 и торий. ' При оценке функции и данных анализов мочи обсуждались факторы, влияющие на выбор соответствующего аналитического метода. Чувствительность и точность метода должны быть адекватны максимально допустимому содержанию изотопа в организме, скорости выде­ ления и частоте взятия образцов; метод должен быть достаточно специфичным или избиратель­ ным во избежание интерференции других радиоизотопов; и, наконец, должны быть определены расходы, связанные во всеми этими факторами, а также скорость и удобства определения. Требование, предъявляемое к чувствительности метода анализа для каждого изотопа определяют путем координации наиболее точных данных скорости выделения с мочой и макси­ мально допустимым содержанием изотопа в организме. Подробно описаны методы анализа мочи, используемые в настоящее время в Управлении по Атомной энергии Соединенного Ко­ р о л е в с т в а . Согласно рассчитанным требованиям к чувствительности с помощью наилучших исполь­ зуемых методов можно обнаружить на основании выделения изотопа с мочой внутреннее за­

ражение порядка 0 , 1 максимально допустимого содержания в организме любого из перечислен­

ных радиоизотопов и порядка 0 , 0 1 максимально допустимого содержания изотопа в организме в случае высокообогащенного урана, трития, полония-210 и цезия-137. ‘

MÉTODOS PARA DETERMINAR RADIONÚCLIDOS EN LA ORINA-APLICADOS POR LA JUNTA DE ENERGIA ATÓMICA DEL REINO UNIDO, Los autores, con la cooperación de químicos analíticos, médicos e higienistas, han estudiado los procedimientos para determinar analíticamente radionúclidos en la orina que actualmente aplica la Junta de Energía Atómica del Reino Unido (UKAEA). Los elementos de mayor importancia en los programas de análisis de'orina de la UKAEA son el plutonio, uranio, tritio y productos de fisión (principalmente estroncio-90 y cesio-137). También se determinan polonio-210, radio-226, protactinio-231, fósforo-32, carbono-14, azufre-35, am ericio-241 y torio. Al evaluar la función y los objetivos del análisis de orina, los autores estudian los factores que ejercen influencia sobre la elección de un método analítico adecuado. 'La sensibilidad y la precisión del método deben guarder cierta relación con la carga corporal máxima admisible, la velocidad de excreción y la frecuencia con que se toman las muestras; el método debe ser suficientemente especifico o selectivo para eliminar toda posibilidad de interferencia debida a otros radionúclidos y, por último, su costo se debe evaluar, en relación con todos estos factores y, también, con la rapidez y el carácter práctico del método. La sensibilidad requerida para cada radionúclido se calcula a partir de la carga corporal máxima admisible aplicando una velocidad representativa de excreción por vfa urinaria, basada en los mejores datos disponibles. Se describen detalladamente los métodos de análisis de orina que acutalmente utiliza la UKAEA. Según los requisitos de sensibilidad calculados, los mejores métodos empleados permiten determinar cuantitativa­ mente, por análisis de orina una contaminación interna igual a un décimo de la carga corporal máxima ad­ misible de cualquiera de los radionúclidos arriba mencionados, y el el caso del uranio fuertemente'enriquecido, RADIONUCLIDES IN URINE 171 del tritio, del polonio-210 y del cesio-137, una contaminación interna cien veces menor que la cargâ corporal máxima admisible. . . .

INTRODUCTION

In work with radionuclides, planning of equipment and operations for maximum safety is supplemented by careful and regular surveillance of the working environment. This provides the basis for a broad assessment of the radiation exposure, from external radiation fields and from internal contamination of the body by radionuclides. It is often necessary to make individual measurements of radiation dose on personnel in addition to the environmental measurements. Personal dosi­ meters can be used to estimate the external radiation dose, but the mea­ surement of the dose resulting from internal contamination is less direct. With some radionuclides, internal contamination can be determined satisfactorily by measurement of the radiation emitted from the body. In other cases, this is not possible at the level required, but an estimate can be made from measurements of the excretion of the radionuclides. The ob­ served excretion rate is compared with reference data from examples in which the level of contamination was known and the urinary excretion has been systematically studied. In the case of most radionuclides, the reference data are rather scanty. For routine surveillance of a large num­ ber of people, even in cases where direct measurement of radiation emitted from the body is technically feasible, the most suitable method, on grounds of economy and convenience, is usually the analysis of urine [1]. The purpose of routine urine analysis is to detect and estimate the level of internal contamination with radionuclides before this level has become significant in any individual case. One of the most important radionuclides with regard to internal conta­ mination is plutonium-239; it is an alpha-emitter of long physical half-life. Insoluble plutonium compounds deposited in the lung only leave it very slowly; soluble compounds, although they enter the bloodstream more rapidly, are then mostly deposited elsewhere in the body, notably in the bone, and sub­ sequently move on even more slowly [2]. Once internal contamination has occurred, prolonged irradiation ensues, regardless of any subsequent pre­ cautions 'against further intake. Although urine analysis is the ultimate mea­ sure of working conditions, it is quite unacceptable as the primary method of environmental monitoring in cases such as plutonium-239. (Deconta­ mination therapy is a separate issue). However, the use of urine analysis for the supplementary assessment of environmental conditions is appropriate for radionuclides which are excreted rapidly, e.g. tritium [3], or which undergo rapid physical decay, e.g. phosphorus-32. In such cases an intake only commits the individual con­ cerned to internal irradiation for a short time and he can be protected from long-term irradiation by guarding against further intakes. In addition to its utility in a programme of routine surveillance of per­ sonnel [4], urine analysis can make an important contribution in the assess­ ment of a recognized accident. Analyses of samples taken soon after the 172 S. JACKSON and ' N. A. TAYLOR accident help to identify the people who have received a significant intake among those presumed to have been at risk. The follow-up investigation can then be concentrated on those shown to be significantly involved. This will probably include measurements of excretion of the radionuclide con­ cerned, in urine and faeces, for as long as is practicable. In many cases this provides the basis for estimation of the size of intake by individuals, and its subsequent change in retention with time, and thus for estimation of the dose of radiation received. The methods used for analysis of urine for radionuclides in the United Kingdom Atomic Energy Authority have been surveyed, in co-operation with analytical, health physics and medical staff throughout the Authority.

2. SPECIAL PRACTICAL FEATURES OF A PROGRAMME OF URINE ANALYSIS

2.1. Estimation of the excretion rate of a radionuclide from the amount found in a sample of urine

It is not easy to define the duration of the period of metabolism repre­ sented by a urine sample. Some constituents of urine are known to be ex­ creted non-uniformly during different parts of the day. In order to eliminate the effects of comparable fluctuations on the results of analysis for a radio­ nuclide, urine is sometimes collected for the whole of a 24-h period. If, however, the body burden of a radionuclide is changing rapidly, a 24-hperiod of collection may be too long. ' The voidings collected during any arbitrary period of time do not re­ present that period of metabolism precisely, because it is most unlikely that the bladder is empty at the beginning of the period and that the last voi­ ding collected occurs exactly at its end. The impact of this discrepancy is reduced if the period of collection is extended. However, in a routine pro­ gramme, prolonged collection is seldom practicable, and few people are capable of sufficient discipline to provide satisfactory samples unless under continuous supervision. Although the volume of urine is obviously the easiest measurable cha­ racteristic, fluctuations in the volume excreted by individuals and between individuals about a mean of approximately 1.4 1 per day are considerable, and make the volume an imprecise measure of the period of metabolic ac­ tivity represented by a urine sample. • The creatinine content of a urine sample may be used to provide an esti­ mate of the period of metabolism which it represents. For an individual, creatinine is subject to smaller fluctuations in excretion rate than the other constituents of urine. However, the creatinine excretion rate varies con­ siderably between individuals, about a mean of approximately 1.7 g per litre. Creatinine content is probably the best index of the representativeness of a urine sample if the mean creatinine excretion rate for the individual is known [5]. The complicated business of determining the excretion rate of a large number of individuals would be hard to justify at present. However, there appears to be a correlation between creatinine excretion rate and RADIONUCLIDES IN URINE 173 muscle mass, and muscle mass can be estimated from simple anthropometric measurements or from measurements of the total body content of potassium [6 ].

2. 2. Standardization of radionuclide determinations Ideally, the analytical recovery of a radionuclide should be estimated for each individual sample. This can be achieved by internal standardization, in which the radionuclide content of a sample is determined before and after the addition of a known quantity of the same radionuclide. An example of this is tiie determination of tritium by liquid scintillation counting (section 5,4). Recovery can be deduced by comparison With that of the stable element, added before analysis. Alternatively, a different radioisotope of the same element distinguishable by characteristically different radioactive decay, may be employed. Strontium-90 recovery is estimated by the recovery of added stable strontium (section 5. 3), and protactinium-231 by the use of protactinium-233 (section 5.6.4). In many cases, the best estimate of recovery which is practically possible depends on the analysis of parallel standards, which can be pre­ pared by adding known amounts of the radionuclide to uncontaminated urine; this is done in the case of plutonium-239 (section 5. 2).

2. 3. Background and blank determinations The limit to the sensitivity of a method of analysis may be set by the magnitude either of the instrument background or of the level found in blank determinations (on the urine of unexposed people). The latter depends on the actual level of activity in normal urine, and also on the level of conta­ mination inevitably introduced in the analytical procedure. When the blank is small compared with the instrument background, fre­ quent determination of the instrument background is essential. The deter­ mination of blanks is only useful for the detection of abnormal contamination in the analytical laboratory. Instrument background determines the limit of sensitivity in the determination of most of the beta-emitters. (The situ­ ation will be different if improved instrumentation leads to a major reduction in background). . ' In other cases, however, e.g. plutonium -239, the instrum ent b ack­ ground may be of small significance compared with the blank. It is then essential to carry out blank determinations concurrently with sample ana­ lysis; determinations of instrument background are required less frequently than when this is the crucial factor. During a period when the blank level is evidently steady, the standard error of the blank estimate can be reduced- by averaging a series of blank determinations.

3. FACTORS INVOLVED IN THE CHOICE OF AN ANALYTICAL METHOD

3.1. Sensitivity and the frequency of routine sampling

Because the quantities of radionuclides to be determined in urine ana­ lysis are often extremely small, the dominant requirement is usually for high sensitivity in the analytical method. 174 S. JACKSON and N. A. TAYLOR

Routine sampling must be frequent enough to ensure that a significant level of contamination is detectable by the chosen method. The relationship between excretion rate and body burden changes as time elapses after the intake of a radionuclide, often rapidly in the case of soluble material. With long sampling intervals the significance of the measured level of urinary excretion is open to wide uncertainty, and the necessarily conservative in­ terpretation may be unduly restrictive. Some radionuclides decay rapidly; some are rapidly excreted; in these cases a programme of frequent samp­ ling is essential. With these provisos, the choice of a more sensitive ana­ lytical method may permit less frequent sampling. (See [45] for further discussion).

3.2. Specificity of selectivity

It is clearly essential that the chosen analytical procedure must elim i­ nate interference by natural radioelements, and by other radionuclides to which a man may be exposed. "Gross" methods, i.e. those which estimate a group of radionuclides, are, inevitably, fraught with a degree of uncertainty. Their proper place appears to be for sorting contaminated personnel from those who are in­ significantly contaminated. Quantitative estimates of internal contamination should preferably be based on specific methods of analysis.

3.3. Precision

The paucity of fundamental data on the metabolism of radionuclides is particularly serious because of the wide margin of uncertainty which is left about the variation between individual people in the rate of excretion. This, and the difficulty in defining the period of metabolic activity repre­ sented by a urine sample (section 2. 1), seriously limits the precision with which the body burden of a radionuclide can be estimated from urine ana­ lysis results. Serious imprecision from these other sources is inescapable, so that high precision in analytical technique is vitiated. This consideration justi­ fies the use of methods such as the fluorimetric method for natural uranium (section 5. 1. 3) which is less precise than other available methods but very much more convenient and economical.

.3.4. Speed and cost of analysis

If alternative analytical techniques offer significant differences in speed, convenience or economy, this may influence the decision when striking-a balance between sensitive analysis and frequent sampling. Speed is im ­ portant if urine analysis is to be used as an indirect means of monitoring the environment, and it may be the over-riding consideration in following-up accidents. Economy may be more important where routine samples are concerned. Savings in skilled man-hours may justify the high capital cost of sophisticated equipment, or may be achieved by the use of simple methods for which comparatively unskilled personnel can be employed. ■ RADIONUCLIDES IN URINE 175

4. SENSITIVITY REQUIRED FOR DIFFERENT RADIONUCLIDES

The radionuclides of major interest in the operations of the United King­ dom Atomic Energy Authority have been considered (Tables I, II, and III). The estimate of sensitivity required has been based on the urinary excretion rate after half the specified sampling interval (which is usually one to three months). Although in most cases the excretion rate is not changing rapidly at this stage, the value chosen is inevitably an approximation; a more de­ tailed consideration of the interdependence of sampling frequency and sen­ sitivity requirement is not possible within the scope of this survey.

TABLE I

SENSITIVITY REQUIRED FROM ANALYTICAL METHODS

. Uranium (section 4. 1)

Enriched uranium Natural uranium (>8.5% U235)

Alpha-activity Alpha-activity C ritica l 4.76 MeV (UiM) C h em ical 4.7 6 MeV (U234) factor 4.1 to 4.6 MeV (U235) to x icity 4 .18 MeV (U2*e> ■ 4.18 MeV (U23s) 4 .1 to 4.6 MeV (U25s)

Soluble Insoluble Soluble Soluble Insoluble com pounds compounds compounds compounds compounds

Critical organ ; Bone Lung ■ Kidney - Bonè Lung

M axim u m permissible • body burden (1) A lph a- 0. 05 jjc 0. 0 2 fic 0 . 01 ¡ic 0. 06 fie 0 .0 2 дс a ctiv ity (2) Mass 15 mg 90 mg 30 mg

Approximate 0|4% 0 .2 % 0 . 4 $ 0 .4 % 0 .2 % • fraction of ( 3 0 - 6 0 d (3 0 -6 0 d (3 0 - 6 0 d body burden sam pling sampling sam pling ex creted in interval) interval) ■interval) urine per day

Approximate am ount excreted in urine when 200 pc/24 h 4 0 p c/2 4 h 40 p c/24 h 200 pc/24 h 4 0 p c/24 h body burden 4 0 j jg /l 200 jig/1 4 0 pg/l is a t the m axim u m permissible level 176 S. JACKSON and N. A. TAYLOR

4.1. Uranium (Table I)

For the soluble case, the data of BERNARD'and STRUXNESS [7] have been used (see JACKSON [8 ]).- For the insoluble case, the figure for the rate of excretion is averaged from the data of FISH [9] and SAXBY et al. [10]. , Enriched uranium (from the diffusion process) containing more than 8.5% of uranium-235 (12 times the concentration present in natural uranium), presents a case critically determined by radiological considerations. (The biggest contribution to the total a-activity is made by uranium-234, because of the enhanced concentration of this isotope. ) A special case, unique among the elements considered bv the International Commission on Radiological Protection (1CRP) is that of inhalation of so­ luble compounds of natural uranium. (Similar considerations apply to the isotopes uranium-238 and uranium-235 alone, or to natural uranium de­ pleted in uranium-235 and uranium-234. ) Inhaled soluble material is rapidly absorbed into the bloodstream; much is rapidly excreted, but there is sig­ nificant retention in the kidney and bone. Although the bone is more critical than the kidney according to radiological criteria, the most limiting factor is chemical toxicity to the kidney, which is therefore the critical organ [10]. With inhalation of insoluble uranium compounds of any isotopic composition, retention of some of the inhaled material causes prolonged irradiation of the lung, which is consequently the critical organ. ■ A s little as 0.01 ng of uranium can be measured by the fluorim etric method, which is usually chosen for the determination of natural uranium. Quantities of natural uranium are usually expressed in mass rather than alpha-activity.

4. 2. Plutonium (Table II) 1

The rate of excretion of plutonium after contamination with soluble com­ pounds is taken from LANGHAM [2]. Langham1 s data have also been used to estimate the rate of excretion resulting from contamination of the lung with insoluble compounds. As sug­ gested by ICRP Committee II (1959), it has been assumed that, in this case, the rate of entry of plutonium into the bloodstream is represented by a bio­ logical half-life of one year [11]. The calculation was made by BEACH [12].

4.3. Strontium (Table II)

The data of BISHOP et al. [13] have been used for the soluble case. For the insoluble case, a similar relationship between body burden and urinary excretion has been assumed, as was found in the case of strontium carbonate by RUNDO and WILLIAMS [14]. •

4.4. Radium

The data for soluble radium are taken from an individual case of acci­ dental inhalation of radium chloride, reported by AUB et al. [15]. The excretion during the period 50 - 90 d after the accident is probably deter- T A B L E П

SENSITIVITY REQUIRED FROM ANALYTICAL METHODS Plutonium, strontium-90, radium-226 and natural thorium

Radionuclide Plutonium-239 Plutonium-241 Strontium-90 R a d iu m -2 2 6 Natural thorium

P rin cip a l A lp h a- A lp h a - B e ta - A lp h a- Alpha-activity ra d ia tio n a c tiv ity . a c ti v i ty a c tiv ity a c tiv ity of decay chain 5 .1 6 M eV of daughter 0 .5 5 M eV . o f d e ca y

(americium- B e ta - ' ch ain 177 URINE IN RADIONUCLIDES 2 4 1 ) a c ti v i ty 5 .4 8 M eV of daughter (yttrium -90) 2 .2 6 M eV

Soluble compounds (bone critical)

Maximum permissible 0 .0 4 цс 0 . 9 ¡ic 2 цс 0 , 1 ¡sc 0 . 01 цс body burden

Approximate 0 ,0 1 % 0 . OVfr 0 . 1 °]o 0 . 01<7o o.oi<7o fraction of body ( 3 0 - 1 0 0 d ( 3 0 - 1 0 0 d ( 3 0 - 2 0 0 d ( 1 0 0 - 2 0 0 d ( 3 0 - 1 0 0 d burden excreted in sam p lin g sam p lin g sam p lin g sam p lin g sam p lin g urine per day in te rv a l) in te rv a l) in te rv a l) . interval) in te rv a l)

Daily urinary excretion when body burden is at 4 pc 9 0 pc 2 0 0 0 pc 10 pc 1 pc = 10 (ig the maximum permissible level 7 - . AKO ad . . TAYLOR A. N. and JACKSON S. - 178

TABLE II (cont. )

Radionuclide Plutonium-239 Plutonium-241 Strontium-90 R a d iu m -2 2 6 Natural thorium

Insoluble compounds (lung critical)

Maximum permissible 0 . 0 2 дс 2 0 fie 0. 8 цс 0 . 0 0 8 fic 0 . 02 цс lung burden •

Approximate A bsorption fraction of lung m akes 0 . 005^° 0 .1 % 0 . 05% burden excreted in bone For natural urine per day c r itic a l thorium a special a fte r curie is defined, Daily urinary ~ 3 0 d; co m p risin g 1 с excretion when s o lu b le. ofThzsz and Jung burden is at 1 p c c a se 8 0 0 p c 4 p c , 1 с o f T h 228 the maximum th e re fo re permissible level a p p lies RADIONUCLIDES IN URINE 179 mined by the very slow rate of departure from bone. Soluble radium pro­ bably enters the circulation very rapidly, and much of it is then deposited in the bone. For the insoluble case, the data reported by MARINELLI et al. have been used [16]. These refer to accidental inhalation of radium sulphate. Paradoxically, there was a higher urinary excretion rate than in the soluble case at the times which can be compared; presumably, the radium in the blood was maintained at a higher level by the continuing slow dissolution of radium sulphate from the lung.

4. 5. Thorium (Table II)

The excretion rate of thorium is very variable, apparently being very sensitive to variations in the form of the contaminating material. Boecker's experiments, on the inhalation of soluble thorium compounds by rats, sug­ gest an excretion rate of 1% after 50 d [17]. However, following ICRP Com­ mittee II [11], it has been assumed that thorium is as persistent in bone as plutonium, and a similar excretion rate has been adopted, namelyO. 01% per day.

4. 6. Tritium (Table III)

Tritium is usually excreted smoothly according to a simple exponential function. . The specified rate of excretion of 6% per day corresponds to a biological half-life of 12 d [11]. This appears to be in the high part of the normal range [3].

4.7. Polonium-210 (Table III)

The best figure for the urinary excretion rate of soluble polonium ap­ pears to be 0.2% per day. Urinary excretion of polonium is fully discussed by TAYLOR [18]. It is difficult to suggest a figure for insoluble material; only a small fraction seems to be absorbed from the lung [19]. -

4.8. Caesium-137 (Table III)

, Human excretion of caesium in urine is 0.5 to 1.5% per day, much more variable than in faeces (RUNDO [20]). Almost all caesium compounds are soluble.

4. 9.; Protactinium-231 (Table III)

Evidence for. the metabolism of protactinium is very scanty. The best figure for soluble compounds, obtained from animal experiments, is 0.2% p er day (E V E [21]) a fter 17 to 6 4 'd, but 0. 01% is used, fo r the same reason as in the thorium case. 8 S JCSN n N A TAYLOR A. N. and JACKSON S. 180 T A B L E III

SENSITIVITY REQUIRED FROM ANALYTICAL METHODS

Tritium, polonium-210, caesium-137 and protactinium-231 (Soluble compounds)

Radionuclide T ritiu m Polonium-210 Caesium-137 Protactinium-231

P rin cip a l Beta-activity Alpha-activity Bèta-activity Alpha-activity ra d ia tio n 1 8 keV 5 . 3 0 M eV 0. 51 MeV . 5. 0 M eV Gamma-activity Alpha-activity

. 0. 6 6 M eV of decay chain

Critical organ Body tissue Spleen ' T o ta l body B one -

M a x im u m permissible 1 0 0 0 ЦС 0. 0 3 fie 3 0 /je 0 , 0 2 jíc body burden

Approximate & ] o 0 . 2 % о. ь °!о 0 . 0 1 °j 0 fraction of (E n tire (E n tire ( 2 0 - 1 0 0 d ( 3 0 - 1 0 0 d body burden excretion curve excretion curve sam p lin g . sam p lin g excreted in can be fitted can be fitted in terv al) in te rv a l) • urine per day with a single, with a single exponential ' exponential fu n ction ) fu n ction )

A m ount excreted in urine when body burden 3 0 j i c f l 60 p c /2 4 h 0.15 jic/24 h 2 p c /2 4 h is a t the (= (M P C )W) m a x im u m permissible l e v e l RADIONUCLIDES. IN URINE- 181

5. URINE ANALYSIS METHODS USED IN THE UNITED KINGDOM ATOMIC ENERGY AUTHORITY

5.1. Uranium

5. 1. 1. General method for enriched uranium [22]

(a) Removal of organic matter and water

Oxidize and concentrate by heating with nitric acid.

(b) Purification of uranium

Add ammonium nitrate and ferric nitrate to the nitric acid solution, as salting-out agents. Extract with diethyl ether continuously for one hour.

(c) Counting -

Evaporate the ethereal extract and ignite on a platinum tray for alpha- scintillation counting.

(d) Limit of sensitivity

1.5 pc/24 h, set by counting system.

(e) Special details of practice of different establishments

Aldermaston Uranium is first co-precipitated with ferric hydroxide + phosphate. Salting-out into ether is by the addition of ammonium nitrate alone; iron is added only in carrier quantity. The source for counting is prepared by electrodeposition under the same conditions as those used for plutonium (vide infra), and the same counting system is also used. This lowers the limit of sensitivity to 0.2 pc/24 h, set by the level of the blank.

Capenhurst Disposable stainless-steel counting trays are now used in place of platinum.

Winfrith . ■ The source is prepared by electrodeposition in the same way as for plu­ tonium, and counted with the same equipment.

5. 1. 2. Method used at Windscale for enriched uranium [23]

(a) Removal of organic matter and water Evaporate with nitric acid. Transfer to crucible. Ignite. 182 S.-JACKSON and N. A. TAJLOR

(b) Purification of uranium . .

Make alkaline (pH 10) with ammonium hydroxide. Collect precipitated phosphates. . Dissolve in concentrated HC1 and potassium iodide. Exchange uranium on to Amberlite IR-400 resin. Elute with IN HC1.

(c) Counting

Evaporate and ignite on a platinum tray for alpha-scintillation counting.

(d) Limit of sensitivity

0.1 pc/24 h, set by the level of the blank.

5. 1.3. General method for natural uranium [24]

Evaporate and ignite a small sample (1 or 2 ml) of urine in a platinum crucible. - . ' . Fuse with 0.6 g of sodium fluoride at 1000°C for 30-70 s. (At some establishments, the sodium fluoride is mixed with sodium bicarbonate, which acts as a flux). Allow to cool. • Measure the yellow-green fluorescence emitted under ultraviolet ir­ radiation, using a photoelectric fluorimeter. ■ The limit of sensitivity is 5 Mg/1. Winfrith achieve a limit of 2 jug/1 by a slight improvement of instrumentation.' ' .

Comments .

- This method is a model of simplicity and speed. Different establish­ ments have encountered problems of fluorescence in blanks attributed to different reagents, which has led to slight differences in practice at the fusion .stage. The required limit of sensitivity corresponding to the presence of the maximum permissible body bürden is 40 jug/1 of urine (Table I) so that the lim it of 5 дg/1 attained in this method does not permit the detection of sig­ nificantly less than one-tenth of the maximum permissible body burden. ‘ ' An improvement in the limit of sensitivity is desirable. It is probable that this could be achieved by re-designing the fluorimeter to a more strin­ gent specification. To add a concentration stage would destroy the conven­ ience of the method for routine analysis of large numbers of samples.

5. 2. Plutonium ’

5. 2. 1. Methods for plutonium-239 using cupferron [25, 26, 27, 28]

Cupferron is used in the isolation of plutonium-239 at most establish­ ments, but there are considerable variations in the procedure. The details are presented in Table IV. , RADIONUCLIDES IN URINE , 183

TABLE IV ' ' '

ANALYTICAL METHODS FOR PLUTONIUM-239 USING CUPFERRON

W indscale Establishment Aldermaston Harwell Win frith and Dounreav

(a) Removal Oxidize with nitric acid. Make alkaline (pH 10)'with

o f organic Take up in hydrochloric acid NH4 OH. Collect precipitate. m a tte r and Ignite at 600°C. Dissolve in w ater hydrochloric acid

(b) Purifica­ Add ferric chloride carrier. Add hydroxylamine. Stand for 1 h. tion of Reduce acidity by adding ammonium hydroxide. Add cupferron. plutonium ■ Stand for 30-45 min to ensure formation of cupferron complex. Extract with chloroform ;

Back-extract Evaporate chloroform. Destroy cupferron by heating

Pu from with H2SO4 and HNO3 chloroform

with 8 M HC1

Extract iron with di-isopropyl ether Evaporate with

HNO3 to rem ove . HC1. Dissolve

in 8 M HNO3. Exchange Pu on to deacidite FF resin (nitrate form). Wash .

with $ M HNO3. Elute with 1 M

HNO3 containing 0 .0 1 M HF

(c ) Source Evaporate HC1. Evaporate HC1. Evaporate HC1. Evaporate HN03.

preparation Add dilute H 2SO4 , Add HNO3 and Add NH4 n itrate, Add HC1 and

hydroxylamine evap orate. NH4 o x a la te and evaporate. and NH4 o x a la te . Heat with H2 S 04. hydroxylamine • Dissolve in HC1.

Neutralize with Neutralize. Add NH4 oxalate

NH4 OH. Acidify Acidify slightly

to 0 .1 N with with H2 S 0 4

H2 S 0 4

Electrodeposit plutonium on a small stainless-steel disc, used as the cathode 184 S. JACKSON and N. A. TAYLOR

TABLE IV (cont).

W indscale Establishment Aldermaston H arw ell W infrith and Dounreay

(d) Counting Transistorized Solid-state, Direct microscope A lpha- system alp h a- energy- counting of scintillation scintillation dependent alpha-particle counter'with counter. d etecto r tracks in sm all ZnS Disposable nuclear emulsion screen $creen prepared plate exposed to for each disc source for 7 d from Sellotape and ZnS powder.

(e) Limit of 0. 04 pc/24 h , 0. 025 pc/24 h 0. 06 pc/24 h 0.02 pc/24 h sensitivity

(a) Removal of organic matter and water

A possible disadvantage of igniting the preparation is the concomitant reduction in solubility as a consequence of sintering. Aldermaston use an entirely wet oxidation, which requires only a small amount of attended time and avoids a transfer stage. 1

(b) Purification of plutonium

Aldermaston first add bisulphate, to remove traces of nitric acid and to avoid the delay involved in standing to wait for complete reduction after the addition of hydroxylamine. In general, a pH of about 1 is used for the cupferron extraction. This appears to give a more reproducible recovery than lower values, together with adequate decontamination from thorium. However, Winfrith consider that pH 0.3 is necessary to prevent the break-through of thorium cupferrate. This is probably important in the case of faecal analysis. . Aldermaston avoid the necessity for evaporation and wet oxidation at this stage by back-extracting plutonium from the chloroform into 8M aqueous HC1. ' It is essential to remove the iron carrier before electroplating. Winfrith achieve this by an ion-exchange technique but the more general practice is to extract the iron into di-isopropyl ether. Dounreay have made preliminary trials of the Winfrith method for iron removal, with a view to bringing it into routine use.

(c) Source preparation

Deposition from a sulphate or oxalate medium is satisfactory. The practice at different establishments differs in detail. RADIONUCLIDES IN URINE 185

To achieve, low blanks, Winfrith prepare their own ammonium oxalate from ammonia and oxalic acid, and recrystallize it.

(d) Counting system ,

The variety of systems used is a result of alternative approaches to the objective of minimum counting background. None of the systems appears to offer decisive advantages over the others.

(e) Limit of sensitivity

The differences in the limits of sensitivity quoted are mainly due to differences in the size of sample analysed.

5. 2. 2. Method used at Springfields for plutonium-239 [29, 30]

The urine is evaporated and wet oxidized. Nitric acid is added to the residue and interfering phosphates are removed by complex formation with concentrated ferric nitrate overnight. Plutonium is reduced to the tetra- valent state with sodium nitrate, and exchanged onto Amberlite IRA-400 resin (chloride form) as [Pu(N 0 3 )s]~. И is recovered by warming at 70°C with hydroxylamine in dilute hydrochloric acid. ' The source is prepared by electrodeposition from an alkaline hypo­ chlorite medium and counted overnight with an alpha-scintillation counter to decide the duration of exposure required with a nuclear emulsion plate. After appropriate plate exposure (up to 7 d) the alpha-tracks are counted with a binocular microscope. The limit of sensitivity is 0.06 pc/24 h.

5. 2. 3, A nalysis f o r p lu to n iu m -241

Plutonium-241 is separated and electrodeposited with plutonium-239. Methods are being developed for counting the low-energy beta-activity of the plated discs by the use qf proportional counters.

5.3. Fission products (Table V)

5. 3. 1. Method for gross beta-activity [31, 32]

The gross beta-activity which is co-precipitated with calcium oxalate is used as a measure of mixed fission products. The principal elements included are the alkaline earths and the rare earths, notably strontium and cerium. Caesium is not precipitated. The sensitivity of the method is greater when the precipitate is counted directly with a thin-end-window G eiger-M ü ller counter, as compared with the a l­ ternative procedure of redissolving the precipitate and counting with a liquid beta-counter. Harwell use a pair of end-window counters surrounded by a shroud of glass-envelope Geiger-Müller tubes arranged in anti-coincidence [32]. Because it is simple and rapid, the method is particularly useful for sorting personnel in a group thought to be involved in an accident. To ensure that the radiation hazard is not under-estimated, it is necessary to assume 186 S. JACKSON and N. A. TAYLOR

TABLE V

ANALYTICAL METHODS FOR FISSION PRODUCTS

Gross beta-activity (Radionuclides co-precipitated Strontium-90 with calcium oxalate on adding (and Caesium-137 Radionuclide ammonium oxalate to urine) strontium-89) (Windscale) (H arw ell) Aldermaston Windscale and and Harwell Dounreay

(a) Remova) of No treatment Evaporate with Add strontium ' Add caesium carrier.

organic matter HNO3 and ignite. carrier. Add Add HNO3 . Evaporate,

and w ater Dissolve in HNO3 . Evaporate. and ignite at 500®C. nitric acid Re-dissolve in Dissolve in HC1

HNO3 or HC1.

(b) Purification Adjust to Add NH4 o x a la te . P recip itate Sr Add silico-tungstic neutral pH. Heat. Neutralize and C a as acid. Collect

Add NH4 o x a la te with NH4OH. phosphates. precipitate by solution. Stand Filter. Wash Extract Ca with centrifuging. 30 min at - precipitate fuming HNOj. Dissolve in NaOH least. Decant with dilute Precipitate Ba • Rem ove anions by • bulk of liquid n h 4o h as chromate adsorption on before and with Ba carrier. d eacid ite FF, resin after Scavenge with column (OH form). centrifuging ferric hydroxide Adsorb Cs.on Zeo- karb 315 resin (NH* form). Elute with 5 M HC1

(c ) Source C o lle ct Dissolve the 1,. For Sr90+ Add perchloric acid preparation precipitate on precipitate in Sr*> and alcohol at 0°C.

paper in 8 M HNOa and Count C o lle ct . perforated adjust volume precipitated p recip itate by

aluminium tray to 1 0 m l o x a la te . centrifuging and by filtration transfer to 2. For Sr*> with suction stainless-steel Re-dissolve. tray • A fter 10 d . or more, count Y*° (equilibrated with Sr90) after precipitating the hydroxide with Y-carrier RADIONUCLIDES IN URINE ' 187

TABLE V (co n t).

Gross beta-activity (Radionuclides со-precipitated Strontium-90 . with calcium oxalate on adding (and . . Caesium-137 Radionuclide ammonium oxalate to urine) strontium-89) (Windscale) (H arw ell) . Aldermaston Windscale and and Harwell D ounreay

(d) Counting End-window Liquid End-window End-window system Geiger-Milller Geiger-Miiller Geiger-Miiller Geiger-Miiller counter. Result counter counter. ' counter. co rre cte d for (Veall type) Precipitate Result corrected for . self-absorption mounted on tray. self-absorption Loss correction and for losses estimated from estimated from S r-carrier Cs-carrier,

(e) Limit of, 1 0 p c /2 4 h 100 pc/24 h 10 p c /2 4 h 15 p c /2 4 h sensitivity (set by counting system)

that all the measured activity is due to strontium-90. Further investigations by urine analysis of individuals who are not cleared by this test should be carried out by measurement of specific radionuclides. Harwell add strontium carrier before oxalate precipitation so that the precipitate can subsequently be analysed specifically for strontium if necessary.

5. 3. 2. Method for strontium '

The Harwell method is presented in Table V [33]. The Windscale method (also used at Dounreay) is slightly different. Pre­ cipitation of the carbonate is used for the final isolation of strontium, and also to concentrate the strontium after the barium chromate precipitation and again after the precipitation of ferric hydroxide [34].

5 .3 .3 . M ethod fo r c a e s iu m -137 .

The Windscale method [35] is described in Table V. At Harwell, gamma-spectrometry is used to measure the caesium-137 content of urine directly. The limit of sensitivity is 110 pc/24 h.

5.4. Tritium

It is general practice in the United Kingdom Atomic Energy Authority to determine tritium by liquid scintillation methods, using coincidence counters .[36]. A KINARDscintillator [37]or a dioxane-based scintillator, is used. This 188 S. JACKSON and N. A. TAYLOR technique makes possible the application of internal standardization. The limit of sensitivity is 0.1 цс to 0.5 ;uc per litre. At Windscale, interfering substances are removed from urine by boiling ' with animal charcoal; at Dounreay the water to be counted is purified by distillation. Calculations [38] based on published data [39] show that thé change in tritium/hydrogen ratio resulting from distillation is less than 3.5%. At Harwell toluene is added before distillation; at Aldermaston the urine is acidified and distilled under reduced pressure.

5.5. Other beta-emitters

The following methods are used at Harwell

5. 5. 1. Carbon-14 is determined by liquid scintillation counting of untreated urine using KINARD scintillator [37]. •

5.5.2. Sulphur-35 is similarly determined, using a dioxane-based scin­ tillator, after wet-oxidizing the urine and evaporating to dryness, to re­ move tritium and carbon- 14. The limit of detection for carbon-14 or sulphur-35 is 0.01 juc/24 h.

5. 5. 3. Phosphorus-32 is determined by counting untreated urine in a low background liquid Geiger-Miiller counter (Veall type). The limit of detection is 500 pc/24 h. Phosphorus-32 present as inorganic phosphate is determined by pre­ cipitating lead phosphate, mounting on a tray and using an end-window Geiger-Miiller counter. The limit of detection is 100 pc/24 h.

5.6. Other alpha-emitters

5. 6. 1. Method for gross alpha-activity (Table VI)

The determination of gross alpha-activity [40] is useful when there is the possibility of exposure to other transuranium elements as well as plu­ tonium. All the transuranium elements are co-precipitated, but uranium, radium and polonium are not. When interpreting the results, it is necessary to assume that the measured alpha-activity is due to plutonium-239 to avoid underestimating the hazard. If it is possible that thorium or a:ctinium is involved, the source should be re-counted after an interval appropriate to permit the decay products, which are not all co-precipitated, to grow into the equilibrium concentrations. i . 5. 6. 2. Method for radium-226

Radium -226 is separated by со,-precipitation with lead sulphate and then with barium chloride, the latter from ethereal hydrochloric acid [33]. Internal contamination with radium may alternatively be estimated from measurements of its daughter, radon, in the breath [41]. T A B L E VI

ANALYTICAL METHODS FOR SOME ALPHA-EMITTERS

Gross alpha-activity (Radionuclides со-precipitated . Polonium-210 R a d iu m -2 2 6 Protactinium-231 Radionuclide with bismuth phosphate (Aldermaston, Harwell and (Harwell) - (H arw ell) and cerium fluoride) W in d scale) AINCIE I UIE 189 URINE IN RADIONUCLIDES (Aldermaston and Harwell)

(a) Removal of Add HNQ3 . Evaporate. Ignite. W et-oxidize with HNO3 Evaporate with HN0 3 and A dd HNO3 . Evaporate. organic matter Dissolve in HNOs, H2 S04. Dissolve in dilute Ignite at 500°C. Dissolve and w a te r HC1 in 1 0 M HC1

(b) Purification Add KCNS. Extract Fe(CNS )3 Cool in ice. Add l/20th volume Add tellurium carrier. Add Extract with into amylalcohol/diethyl of concentrated H2 S 0 4 . Add Na-hypophosphite. Boil gently. di-isopropylketone. ether. Add NH4 O H , Add lead carrier. Stand 15 min Collect precipitate by Evaporate the extract and, H2 S03. Add phosphoric acid. at 0°C. Collect precipitate filtration. Dissolve in ig n ite . . Add bismuth carrier. Collect by centrifuging. Dissolve in brominated HC1. Warm. Add precipitate by centrifuging. HCl/ether. Cool in ice. hydrazine. Boil gently. Dissolve in dilute HC1. Add Add barium carrier. Collect Filter off the tellurium and cerium carrier. Add HF. precipitate by centrifuging d iscard it Collect precipitate by centrifuging • 9 - . AKO ad N A TAYLOR A. N. and- JACKSON S. - 190

TABLE VI (cont. )

Gross alpha-activity (Radionuclides со-precipitated ' Polonium-210 R a d iu m -2 2 6 Protactinium-231 Radionuclide with bismuth phosphate (Aldermaston,. Harwell and (H arw ell) (H a rw e ll) and cerium fluoride) W in d scale) (Aldermaston and Harwell)

( c ) S o u rce Transfer the precipitate to Transfer the precipitate to a Add HC1. Allow 5 h or more Transfer to a counting tray preparation a platinum tray platinum tray. Add a few for spontaneous deposition of with the aid of HNOg. drops of dilute H^O^. Po on a silver disc. Wash Evaporate . . E v ap o rate and dry

■ (d) Counting Alpha-scintillation counter Alpha-scintillation counter Alpha-scintillation counter Alpha-scintillation counter sy stem

(e) Limit of 0.1 pc/24 h . 0,1 pc/24 h 0.1 pc/24 h 0.1 .pc/24 h ■ sensitivity RADIONUCLIDESíIN URINE 191

5. 6. 3. Method for polonium-210 (Table VI)

After co-precipitation of the polonium with tellurium as carrier, and dissolving in brominated hydrochloric acid, the tellurium is removed by reduction with hydrazine. The polonium will then deposit spontaneously on a silver disc [42]. ,

5. 6. 4. Method for protactinium-231 (Table VI)

At Windscale, a complex chloride ion-exchange step is used in addition to the procedures described in Table VI, as used at Harwell, and protactinium-233 tracer is added to determine recovery.

5. 6. 5. Method for natural thorium [43]

Hitherto there has been no regular demand for urine analysis for natural thorium in the United Kingdom Atomic Energy Authority. The fol­ lowing method based on the work of PERKINS and KALKWARF [44] has been used occasionally at Springfields. Co-precipitate thorium with lanthanum fluoride. Dissolve in nitric acid. Extract with thenoyltrifluoroacetone. Back-extract into nitric acid, react with morin and estimate by spectrophotometry.

6. APPRAISAL

Further improvement in the precision of estimation of internal con­ tamination from urme analysis results depends primarily on the emergence of better data about the metabolism and excretion of radionuclides. The analytical methods described here all achieve high levels of recovery, and well-defined standards of specificity. According to the cri­ teria developed in section 4, all the methods are capable of measuring in­ ternal contamination at a level of one-tenth of the maximum permissible body burden. With enriched uranium, tritium, polonium-210 or caesium-137 one-hundredth of the maximum permissible body burden can be quantitatively detected. Most of the methods described here are readily applicable to the analy­ sis of faeces, which can be a valuable source of information. Apart from the need for slightly different preliminaries, the only complicating factor is the presence of significant amounts of natural alpha-emitters in faeces, notably thorium-228 and its decay products. It has been pointed out in section 3. 1 that analytical sensitivity and sampling frequency are inter-related. If a very sensitive analytical method is not available, high frequency of sampling is necessary. However, in the practically important case of plutonium-239, measurement of internal con­ tamination with one-hundredth of the maximum permissible body burden is not easy even with a monthly sampling programme, in spite of notably suc­ cessful efforts at increasing the analytical sensitivity. In this situation the problem might be eased by collecting larger samples of urine representing periods greater than 24 h, and containing proportionately more plutonium. 192 S. JACKSON and N. A. TAYLOR

With some radionuclides, e. g. plutonium and uranium, urinary excre­ tion fluctuates widely from day to day, to an extent unlikely to be accounted for by the combined effects of analytical error and the uncertainty in de­ fining the metabolic period represented by each sample. There is; as yet, no biological interpretation for these fluctuations, and the only practicable course in utilizing the results of a series of urine analyses is to use them to derive an averaged excretion curve. Since sampling data must, there­ fore, be averaged during interpretation, a comparable result would be achieved by pooling aliquots of a series of samples and carrying out one analysis only on the pooled aliquots. The remaining major portions of the individual samples would be available for individual analysis if this was re­ quired after consideration of the result from the pooled aliquots. It is prob­ able that this procedure would be more economical because of a reduction in the total number of analyses required. 'Some of the possibilities mentioned here are considered in greater de­ tail by DOLPHIN, JACKSON and LISTER [45].

ACKNOWLEDGEMENT '

The authors are most grateful to all those concerned with urine analysis throughout the United Kingdom Atomic Energy Authority for their willing co-operation and invaluable advice in the course of this survey.

REFERENCES . -

[1] -LISTER, B. A. J. , "The problems and methods of sample assay”, Diagnosis and Treatment of Radioactive Poisoning, IAEA, Vienna (1963) 23-40.

[2] LANGHAM, W. H. , Brit. J. R ad io l., Suppl. N0 . 7 (1957) 95-113. [3] W YLIE, K. F. , BIGLER, W. A. and GROVE, G. R. , Hlth Phys. 9 (1 9 6 3 ) 9 1 1 -4 . [4] SHERWOOD, R. J. and SPOOR, N. L. , AERE H P/R 2581 (d el) (1 9 5 8 ). [5] JONES, O. ,. WSL/417 (1952).

. [ 6 ] BARTER, J. and FORBES, G. B. , Ann. N. Y. Acad. Sci. 110 (1963) 264-70. [71 BERNARD, S. R. and STRUXNESS, E. G ., ORNL-2304 (1957).

[8 ] JACKSON, S. , AHSB(RP)-R15 (1962). [9] FISH, B. , in: Inhaled Particles and Vapours (DAVIES, C. N. , Ed.), Pergamon Press (1961) 151-65. [10] SAXBY, W. N. , TAYLOR, N. A ., GARLAND, J., RUNDO, J. and NEWTON, D. , "A case of inhalation of enriched uranium dust", these Proceedings. [11] Recommendations of the International Commission on Radiological Protection, Report of Committee II on permissible dose for internal radiation, Pergamon Press (1959). [12] BEACH, S. A ., personal communication (1963). [13] BISHOP, Margaret, HARRISON, G. E. , RAYMOND, W. H. A ., SUTTON, Alice and RUNDO, J. , Int. J. Rad. Biol. 2 (1960) 125-42. [14] -RUNDO, J. and WILLIAMS, Katharine, Brit. J. Radiol. 34 (1961) 734-40. [15] AUB, J.C. , EVANS, R. D. , GALLAGHER, D. M. and'TIBBETTS, Dorothy, M. Ann. internal Med. 11 (1938) 1443-63. [16] MARINELLI, L.D. , NORRIS, W. P. , GUSTAFSON, P. F. and SPECKMAN, T. W. , Radiology 61 (1953) 9 0 3 - 1 5 . . [ 1 7 ]. BOECKER, B. B. , U R -6 0 5 (1 9 6 2 ) 183. [18] TAYLOR, N. A. , Human Excretion Data for Polonium-210 and the Estimation of Body Content and Radi­

ation Dose, to be published. 1 [19] FOREMAN, H ., MOSS, W. and EU ST LE R, B .C ., Amer. J. Roentgenol. _79 (1958) 1071. RADIONUCLIDES IN URINE 193

[20] RUNDO, I., Brit. I. Radiol. 37 (1963) 108. [21] EVE, I. S. , unpublished information (1958). [22] U. К. A. E. A. Report (Capenhurst) IGO-AM/CA-79 (1957). [23] DALTON, J. , unpublished work. [24] PRICE, G .R., SERRETTI, R. G. and SCHWARTZ, S., Anal. Chem. 25(1953) 322. [25] DALTON, J. , PG Report 284(W) (1962). [26] BAINS, M. E. D. , AEEW-R292 (1963), [27] SANDALLS, F. J. and MORGAN, A. , AERE R-4391 (1964). [28] BROOKES, I. R. , to b e published.

[29] SANDERS, S. M. and LEIDT, Sarah, Hlth Phys. 6 (1961) 189-97. [30] JACOBSEN, W.R. , ANL-6637 (1961) 33. " [31] U. К. A. E. A. Report WSL/M-432 (1953). [32] THORNETT, W. H. , AERE-R3172 (1961). [33] BROOKS, R. O. R. , A ERE-AM 60 (1 9 6 0 ). [34] U.K. A. E. A. PG Report 99(W) (1960). [35] U. К. A. E. A. PG Report 98(W ) (1960). [3 6 ] U. K . A. E. A. PG Report 1 6 2 (W ) (1 9 6 0 ). [37] KINARD, F .E ., Rev. sci. Instrum. 28(1957 ) 293 - 94. [38] RILEY, C. J. and BROOKS, H. , unpublished work (1963). [39] AVINUR, P. and NIR, A., Nature 188 (1960) 652. ' [4 0 ] JENKINS, E. N. and SNEDDON, G. W. , AERE C /R 139 9 (1 9 5 4 ). . [41] GROVE, W. P. and C LA CK , B. N. , Brit. J. R a d io l., Suppl. No. 7 (1 9 5 6 ) 1 2 0 -3 . [42] U. К. A. E. A. Report IGO-AM/W-167 (1958). [43] HELMAN, A. H. , U. К. A. E. A. Report P G -1 7 8 (S ) (1 9 6 1 ). [44] PERKINS, R. W. and KALKWARF, D. R. , Anal. Chem. 28 (1956) 1989. ' [45] DOLPHIN, G. W. , JACKSON, S. and LISTER, B. A. J. , "Interpretation of bio-assay data", these Proceedings.

DISCUSSION ■

H. WYKER: In section 5.1.3 it is stated that an improvement is desirable in the sensitivity of uranium determination in urine as the limit reached in the United Kingdom is about 2 ,ug/l. In our establishment, the N. V. KEMA at Arnhem, Netherlands, we are working on the development of a suspension reactor. For our experi­ ments we use hundreds of kilograms of fine crystalline particles, between 2 and 6 цт diam., of a mixture of 15% U02 and 85% ThC>2 . We wondered if thorium build-up in the body through chronic inhalation of these particles could be followed by measuring uranium excretion in the urine. This is only feasible if uranium concentrations much lower than the lim it mentioned above can be measured, and we believe we have reached a limit of 0.01 Mg/1- Our method is not new. It is a fluorimetric method described in the Report of a Joint WHO/FAO Expert Committee.* From each urine specimen we take 3 samples of 0.1 ml, which are/pro­ cessed and measured independently. The time required is about 15 min per urine sample. In the region of 10-12to 10-11 g U, corresponding to concen­ trations of 0.01 to 0.1 ng/l, the standard deviation as determined from about 170 tests is about 0.002 цg/l. We think that, apart from high stability of the fluorimeter, the most important requirements to reach this figure are very great accuracy and

* Methods of Radiochemical Analysis, World Health Organization, Technical Reports Series, No. 173, Geneva (1959). 194 S. JACKSON and N. A. TAYLOR very stringent limits in the preparation of the samples. A sample, for instance, for which the required time of heating appears to be 1 s outside the set limits, has to be discarded. As we cannot help wondering why nobody before us has obtained similar results, we shall not publish them until they are firmly checked. A scheme has been worked out to have a number of urine samples measured inde­ pendently by us and by the health physics group of the RCN in Petten, North Holland, which also uses a very sensitive method, though with a much more laborious preparation technique. Only if there is agreement will we place absolute reliance on our own results and publish the details which, in our opinion, might have led to the increased sensitivity. RECENT RADIOCHEMICAL PROCEDURES FOR BIO-ASSAY STUDIES AT TROMBAY

P.R. KAMATH, I.S.B H AT, KAMALA RUDRAN, .M . A.R. IYENGAR, . ELIZABETH KOSHY, URMILA S. WAINGANKAR AND VASANTI S. KHANOLKAR HEALTH PHYSICS DIVISION, ATOMIC ENERGY ESTABLISHMENT TROMBAY, BOMBAY, INDIA

Abstract — Résumé — Аннотация — Resumen

RECENT RADIOCHEMICAL PROCEDURES FOR BIO-ASSAY STUDIES AT TROMBAY. The bio-assay labora­ tory at Trombay carries out nearly 2000 analyses of urine samples annually for different radionuclides. The most abundant isotopes o f interest are T h 232, T h 228, Ra226, Ra224 and Ra228, U n atu ral, fission products and Pu239. T h e others are C o 60, I 131, P32, C s137, H3, Sr90, e t c . Analyses of a large number of samples containing low levels of activity with speed calls for development of special techniques when high specificity with good yields are desired. Urine presents a very complex matrix with a large amount of complexing organic material as well as variable salt content. Wet ashing procedures are time-consuming and cumbersome where samples of about 1 to 1 . 5 1 are required to be destroyed. The paper presents some very recent methods developed in the laboratory and details of tracer studies involved in the development of these methods.

Ra228: Measurement of mesothorium in urine is made after separation of Ac 228 grown in equilibrium with it. Radium present in urine is со-precipitated with BaS04 and A c 228 allowed to grow. The precipitate is dissolved in perchloric acid after the addition of Pb, Bi and La carriers. The insoluble sulphates are precipi­ tated with diluted H?S04. Ac 228 in the supernate is carried on LaF3 and counted for beta activity and checked for decay. Ra228 is calculated from observed Ac 228 activity after correcting for decay and chemical recovery. Thorium: The method describes techniques developed for the со-precipitation of thorium with calcium oxalate without wet ashing'urine. A study of exact experimental conditions for quantitative precipitation of thorium is reported. The paper also gives tracer experiments carried out for study of interference from phosphates and calcium in the final extraction of thorium with TTA. Thorium-232 is estimated spectrophotometrically and thorium-228 in the separated thorium fraction by separation of Ra224 after equilibration. .

Among other studies reported are the determinations of Cs137, P32, Co 60 and I131. Urine is oxidized by boiling with nitric acid and hydrogen peroxide and proceeded with for the estimation of these elements; Cs137 is separated by absorption on ammonium phosphomolybdate; a study is made of the different methods for the recovery of Cs137 from AMP; P32 is precipitated as MgNH4P 0 4 in the presence of EDTA as masking agent;

C o 60 is recovered as chloro complex by ion exchange techniques; I131 is separated from urine with silver chloride by bath extraction techniques. The paper reports progress of work in the bio-assay laboratory from 1958 to date.

■ MÉTHODES RADIOCHIMIQUES D'ANALYSE BIOLOGIQUE EMPLOYÉES A TROMBAY.. Le laboratoire d'analyses biologiques de Trombay procède chaque année au dosage de différents radionucléides dans près de 2000 échantillons d'urine. Parmi les éléments intéressants, ceux que l'on rencontre le plus fréquemment sont le thorium 232 et 228, le radium 226, 224 et 228, l'uranium- naturel, les produits de fission et le plutonium 239. Les autres sont le cobalt 60, l'iode 131, le phosphore 32, le césium 137, le tritium, le strontium 90, etc. • ' Pour analyser rapidement un grand nombre d'échantillons faiblement actifs, il faut mettre au point des techniques spéciales si l’on veut une haute spécificité et de bons résultats. L'urine constitue une matrice très complexe, car elle contient une grande quantité de complexants organiques et sa teneur en sels est variable. Lorsqu'il s'agit de détruire des échantillons de 1 à 1, 5 1, les procédés d’incinération par voie humide s’avèrent longs et difficiles à mettre en œuvre. Le mémoire expose quelques-unes des méthodes que le laboratoire a mises au point tout récemment, ainsi que les détails d'études à l'aide de radio-indicateurs faites pour ces mises au point.

195 196 P.R. KAMATH et al.

228Ra: On dose le mésothorium dans l'urine après avoir séparé le 228Ac formé en équilibre avec lui.

On fait coprécipiter le radium présent dans l'urine avec BaS0 4 et on laisse se former 228Ac. Le précipité est dissous dans i’acide perchlorique après addition d'entraîneurs (Pb, Bi et La). Les sulfates insolubles sont préci­

pités avec H2S04 dilué. 228Ac surnageant est entraîné sur LaF3 ; on en mesure 1*activité bêta et en contrôle la décroissance. D'après la teneur en 228Ac observée, que l'on corrige pour tenir compte de la décroissance

et de la récupération chimique, on calcule la teneur en 228 Ra. . Thorium: Le mémoire décrit les méthodes que l’on a mises au point pour coprécipiter le thorium avec l'oxalate de calcium, sans incinération de l’urine par voie humide. Il fait état d’une étude sur les conditions exactes d'expérimentation nécessaires pour la précipitation quantitative du thorium. Il y est aussi question d'expériences, à l’aide de radioindicateurs, sur l'interférence due aux phosphates et au calcium dans l’extraction finale du thorium au moyen de TTA. On mesure le thorium 232 par spectrophotométrie; quant au thorium 228 contenu dans la fraction de thorium séparée, on l'estime par séparation de 224Ra après équilibrage.

Au nombre des autres études mentionnées figure la détermination de 1 3 7 C s, 3 2 P, 60Co et 131I¿ On oxyde l'urine par ébullition avec de l'acide nitrique et de l'eau oxygénée, et on la traite ensuite en vue de la détermination de ces éléments. On sépare 137Cs par absorption sur du phosphomolybdate d'ammonium.

Les auteurs indiquent les différentes méthodes permettant de récupérer 137 Cs contenu dans le phosphomolybdate. On précipite 32P sous forme de MgNH4PQ en présence d'EDTA intervenant comme agent de blocage. On récupère 60Со par échange d’ions sous forme de complexe chloro. On sépare 1311 de l'urine au moyen de chlorure d’argent par des méthodes d'extraction en discontinu. Le mémoire signale les travaux accomplis au laboratoire d'analyses biologiques depuis 1958.

НЕДАВНО РАЗРАБОТАННЫЕ В ТРОМБЕЕ РАДИОХИМИЧЕСКИЕ МЕТОДЫ ДЛЯ БИО­ ЛОГИЧЕСКИХ ИССЛЕДОВАНИЙ. В лаборатории биоанализа в Тромбее ежегодно проводится анализ 2000 проб мочи на содержание различных радиоизотопов. Наиболее часто встречаются

изотопы Th232, Th228, На2 2 6 , R a 2 2 4 и R a228, прирдный уран, продукты деления и Pu239. Кроме т о г о , и с с л е д у ю т с я С о 6 0 , J 131, Р 32, C s 137, H 3 , S r^ o и д р . .Быстрое проведение анализов больших количеств проб, содержащих низкие уровни актив­ ности, требует разработки специальных методов, если желательно получить в ы с о к у ю сп е ц и ­ фичность с хорошим выходом. Моча представляет собой очень.сложную матрицу с большим количеством сложных органических веществ и изменяющимся содержанием солей. Процедура получения влажного пепла требует большой затраты времени и является трудной, если необхо­ димо разрушить пробы объемом 1 —1,5 л. В докладе представлены некоторые новейшие разработанные в лаборатории методы и под­ робные данные об излучении индикаторов, используемых при разработке этих методов. Радий-228: измерение мезотория в моче производится после отделения Ас 2 2 8 , о б р а з о ­ вавшегося в равновесии с ним. Радий, находящийся в моче, соосажден с BaS04 и оставлен

для образования Ас 2 2 8 . Осадок растворяется в хлорной кислоте после добавления носителей Pb, Bi и La. Нерастворимые сульфаты осаждаются разведенной серной кислотой. Ас 2 2 8 в надосадочной жидкости переносится на L aF 3 и подсчитывается на бета-активность, а затем

проверяется на'распад. Ra 2 2 8 рассчитывается из активности Ас 2 2 8 после поправки на распад и химическое восстановление. Торий: описывается метод, разработанный для соосаждения тория со щавелевокислым кальцием без влажного сжигания мочи. Сообщается об изучении точных экспериментальных условий количественного осаждения тория. Сообщается об экспериментах с индикаторами, проводившихся для изучения интерференции от фосфатов и кальция в конечном экстракте тория с Т Т А . Торий-232 определяется спектрофотометрическим способом, а торий-228 в отдельной фракции тория путем отделения R a-224 после установления равновесия.

Кроме того, сообщается об определении C s137, Р 32, Со 6 0 и J 131. М оча о к и сл я е т с я п утем кипячения с азотной кислотой и перекисью водорода, и после этого исследуется содержание

этих элементов. Cs 1 3 7 отделяется абсорбцией на фосфомолибдате аммиака. Изучаются раз­ личные методы восстановления C s 1 3 7 и з А М Р . Р 3 2 осаждается в виде MgNH 4 P 0 4 в п р и су т ­

ствии EDTA в качестве действующего агента. Со 6 0 восстанавливается в виде хлорокомплекса

методами ионообмена. J 1 3 1 выделяется из мочи хлористым серебром методами экстракции в в а н н е . В докладе сообщается об успехах работы лаборатории в области биоанализа в период с 1958 года и до настоящего времени. PROCEDURES FO R B IO ASSAY 197

REGIENTES MÉTODOS RADIOQUÍMICOS APLICADOS AL ANÁLISIS BIOLÓGICO EN TROMBAY. El laboratorio de análisis biológicos de Trombay efectúa anualmente cerca de dos mil análisis de orina para

determinar diferentes radionúclidos, siendo los isótopos de mayor interés el 2 3 2 T h , 2 2 8 T h , 226 R a, 224 Ra, 2 2 8 Ra,

U natural, productos de fisión y 239Pu, asfcomo e l 6 0 C o , 32p> i3 7 ç S( эодг, e t c . . El análisis rápido de un gran número de muestras de baja actividad exige el perfeccionamiento de técnicas especiales, sobre todo cuando se desea un alto grado de especificidad, unido a un elevado rendimiento. La' orina presenta una matriz muy complicada, con una elevada proporción de sustancias orgánicas complejantes y contenido variable de sales. Los procedimientos de calcinación por vía húmeda exigen un tiempo con­

siderable y resultan engorrosos cuando se han de destruir muestras de aproximadamente 1 a 1 , 5 1 . El trabajo deseribe algunos métodos recientemente desarrollados en el laboratorio de los autores y propor­ ciona detalles de estudios sobre indicadores que han intervenido en el desarrollo de aquellos métodos. Radio-228: Se mide el mesotorio contenido en la orina después de separar el 228Ac en equilibrio radi­ activo. El radio presente en la orina se coprecipita con BaS04, teniéndose en cuenta que sigue formándose

228Ac. El precipitado se disuelve en ácido perclórico después de añadir como portadores Pb, Bi y La. Los

sulfatos insolubles se precipitan con H2 S 0 4 diluido y el 228Ac que permanece en solución, después de ser

arrastrado con F3 La, se somete a recuento de actividad beta y se comprueba la disminución de actividad.

La cantidad de 224la se calcula a partir de la actividad observada en el 228 Ac después de introducir la oportuna corrección para tener en cuenta la desintegración y el rendimiento de la recuperación química. Torio: El trabajo describe la técnica puesta a punto para la coprecipitación del torio con oxalato cálcico sin necesidad de calcinación por vía húmeda de la orina e informa acerca de un estudio de las condiciones experimentales correctas para la precipitación cuantitativa del torio. El trabajo menciona asimismo experi­ mentos con indicadores llevados a cabo para estudiar las interferencias causadas por los fosfatos y el calcio en la extracción final del torio con tenoiltrifiuoroacetona. La valoración del torio-232 se efectúa por espectrofotometría y la del torio-228 - en la fracción del torio remanente- por separación del 224Ra después de alcanzar el equilibrio.

Otros estudios citados en la memoria incluyen la determinación del l37C s, 3 2 P, 60C o y 131I. La orina se oxida hirviéndola con ácido nítrico y peróxido de hidrógeno antes de proceder a la valoración de estos elementos. El 137Cs se separa por absorción en fosfomolibdato amónico y se realiza un estudio de los diferentes métodos para la recuperación de aquel elemento del fosfomolibdato amónico. E132P se precipita bajo forma de MgNH4P04 en presencia de etilendiaminotetraacetato como agente de quelación. Él 60Co se recupera como cloro-complejo por medio de técnicas de intercambio iónico y el 13iI se separa de la orina con cloruro de plata mediante técnicas de extracción discontinua. La memoria describe la actividad desplegada por el laboratorio de biología desde 1958 hasta la fecha.

The assessment of body burden through excretion analysis is indirect and involves several uncertain factors in actual .computation. The whole- body monitor gives a direct measure of the total-body radioactivity wherever the technique is applicable. Both the indirect and the direct techniques are complementary in health physics monitoring and have characteristic lim i­ tations and problems in their application. For the calculation of body burden from urinary results, it is necessary to have an exact knowledge of the average rate of elimination in an individual as w ell as the total output in urine in relation to the total-body radioactivity. Excretion patterns for several radionuclides have been obtained.through human and animal experimentation [1-5), following an acute exposure and after the activity had deposited in the body. It is necessary to collate and analyse this vast amount of information and evolve broad excretion patterns of different elements for the various modes of internal contamination. It has been possible to apply experimental information obtained from such studies to arrive at total intake and calculate exposure dose [6]. The whole-body monitor is a gamma spectrometer with its problems of geometry, high background, low sensitivity and unfolding of complex 198 P. R. КАМА-TH et al. spectra. The sensitivity of the instrument has been reported as being lOOOpc for certain radioisotopes [7], which is about one-tenth of the maximum p er­ missible body burden (MPBB) for the most toxic radioelements. The instru­ ment does not detect alpha-emitters, which form the most hazardous internal contamination, and soft beta emitters. The problems of calibration for different geometries are not completely solved and analyses of spectra in mixed contaminations could be very time consuming. In situations where the instrument can be used, it is quite a facile tool and can measure the actual body burden with high accuracy. Urine analysis is a very useful technique for the detection of exposure as well as for routine monitoring. The specific radionuclides can be separated from complex contaminations with a remarkable degree of purity and high efficiency. The method is applicable for the detection of all iso­ topes, the limit for the detection being as low as one-millionth of MPBB in most cases. One of the main advantages in its application is that a large number of urine samples can be processed at a time without disturbing the staff from their work schedule. '

TABLE I

MAXIMUM PERMISSIBLE URINARY EXCRETION LEVELS FOR RADIOTOXINS

M axim um Calculated maximum permissible Daily excretion permissible urinary Radioactivity body burden rate in urine excretion level (ЦС) . (%) ( 1 tolerance) (d p m /2 4 h)

Gross alpha Ranges from 0.01 jjc 0 .Ó 1 4 . 4 for natural thorium to 0 . 05 цс for Am

Gross fission 2 (strontium-90) 0 .0 5 2 0 0 0 products

Plutonium-239 0 .0 4 0 . 0 1 8 . 8

Polonium-210 0 .0 3 0 . 0 2 1 3 .3 2

R ad iu m -226 0 . 1 0 . 0 1 2 2 . 2

N atural 0. 01 (89 mg) 0 . 0 1 4 . 4 thorium ( 8 . 9 (jg)

N atural 0 .0 1 5 g V ariab le 50 fig/1 uranium

Phosphorus- 32 6 0 .1 7 2 2 800

Io d in e -131 0 . 1 0 . 1 1 5 5 4

C o b a lt-60 1 0 4 .7 1 .0 4 5 X 10 6

Tritium 1 0 3 5 .3 28 iic/1

Caesium-137 30 0 .1 9 ■ 6 6 600 PROCEDURES FOR BIO-ASSAY 199

The bio-assay laboratory at Trombay was started in 1957 for the de­ tection and measurement of internal contamination through urine analysis. A 16-h overnight sample is collected and analysed for specific radionuclides by methods given in the AEET manual [8]. The results of the analyses are reported in terms of body tolerances calculated on the basis of MPBB [9] and release rates of the slow excretion component commonly adopted for the element [10, 11] (Table I). The size of the sample for actual analysis, however, depends on the detection requirements set for control monitoring. For occupational monitoring it is adequate to set the control limits for urinary contamination, say, at 1/10 tolerance le^el. With a defined limit, the size of the sample is governed by the detection limit of the instruments. Figure 1 gives the progress of work load of the laboratory for urine analyses since 1958. Table II gives a break-up for the different radio­ nuclides tested. "

TABLE II

BREAK-UP OF URINALYSIS SAMPLES DURING 1958-63 PERIOD

Serial No. E lem ent 1958 19 5 9 1960 1961 1962 1963

1. U ranium 58 2 1 5 512 760 834 521

2 . Thorium 205 180 312 565 485 468

3 . Plutonium 2 0 2 0 60 15 64 ' 27

4 . Fission product 18 2 0 130 105 149 146 (gross beta)

5 . Others 8 13 62 136 189 2 2 3 *

Detailed break-up of other samples (1963):

of sam ples Analysed for No. of samples Analysed for

57 Gross alpha . 1 Actinium- 227

65 Radium- 226 1 Polonium- 210

228 2 Europium-152 41 Tritium 23 Phosphorus-32 32 Io d in e-131 . 1 Sod iu m -24

The paper presents some of the more recent methods developed in the laboratory which, in some respect, improve upon existing methods both in speed of manipulation and ease of separation. Detailed procedures for thorium (nat), radium-228, radiocobalt, radioiodine and radiocaesium are given in an Appendix to the paper. Development work for these was necessi­ tated by an excessive work load on some (Th232), lack of suitable procedures for others (Ra228, Co60) and the attempt to minimize delays in analytical output (I131). This is in fact part of our search for rapid and sim ple radio­ chemical methods for low-level measurements. Excretion studies have shown that urinary and faecal eliminations are influenced by the soluble or insoluble forms of the element, the soluble appearing largely in urine. These considerations indicate intuitively the 200 P.R. KAMATH et al.

YEAR

F ig -1

■ ' Work load of bio-assay laboratory (1958-63) appropriateness of external spiking with soluble activity adopted for the standardization of methods. The efficiency of radiochemical separation was checked by analysis of spiked urine samples, containing other activities likely to occur or similar in chemical behaviour including a gross spectrum of elements from a mixed fission-product solution. For the estimation of Ra228, which was measured by counting beta-emitting Ac228, the purity was confirmed by half-life determination. The efficiencies were determined both by chemical and spike recoveries.

DEVELOPMENT OF NEW METHODS

Thorium (nat)

The original method for the determination of natural thorium consisted in wet ashing urine and co-precipitating thorium as the fluoride on lanthanum carrier. Thorium was recovered subsequently in carrier-free state by thenoyl trifluoro acetone (TTA) extraction and estimated as the thorium- thoronol complex by spectrophotometry. The estimation took about three days and the recovery was 65%. PROCEDURES FOR BIO-ASSAY 201

16-h urine + 300 mg Ca boil, + saturated oxalic acid

1 Precipitate Supernatant Dissolve in HC1 and Reserve for Ra recovery precipitate with 3% oxalic acid 1* i Precipitate Supernatant and washings

Oxidize with HNO3 + HC1 04, take in 0. IN Щ Оэ, extract f with T T A in benzene

1 Organic layer Aqueous residue

Back extract with 2N HNO 3

' 1 Organic layer Aqueous layer

Reserve for recovery of TTA Fume with HClOj, take in . dilute НСЮ 4 , add thoronol, measure absorbency at .r)4f)nm

F ig . 2

Analytical sketch for thorium (nat) The revised method is outlined in Fig. 2. Thorium present in urine is carried on calcium oxalate, precipitated at a pH of about 1 and is separated from adsorbed phosphate by a second precipitation. The oxalate is destroyed with perchloric acid, thorium separated by TTA extraction and then taken up for spectrophotometric measurements. The suitability of the method was checked by using urine spiked with Th234 as nitrate as well as citrate. The. overall recovery for the method was 75% and the estimation could be com­ pleted in a day and a half. Urinary calcium could be used for со-separating thorium in principle, provided calcium content was uniform and adequate. Addition of external calcium removed dependence on vagaries of urinary calcium content, which at tim es gave very low values (100 m g). Experiments with Th234 in solutions containing 200-600 mg Cá demonstrated.that TTA extraction efficiency was • not affected by Ca in this range. Earlier experiments using 100-200 mg Ca in 500 m l urine and precipitating the oxalate showed that tracer Th234 was quantitatively recovered in the precipitate. However, when the precipitate was taken up for TTA extraction after treatment, the recoveries were un­ predictable and ranged from 20% to 70%. This was suspected due to the presence of phosphate and experiments were carried out to study phosphate interference. - . Estimation of phosphates in urine and aqueous residue showed that about 0.4-1 mg P equivalent was recovered in the aqueous residue from TTA extraction step: '

P in 24-h urine sample P in aqueous residue ' (mg) (mg) 285 0.43 . 205 0.64 295 1.02

Varying amounts of P as phosphate were introduced in a solution con­ taining 300 mg Ca-carrier and tracer Th234. TTA extraction of Th was 202 P.R. KAMATH et al. carried out at pH 1. The aqueous residue was taken up and Th234 estimated in it by со-precipitating with BiP04. The results showed that there was considerable interference in extraction when the P content was greater than 0. 25 - 0. 3 mg.

P content in solution Th234 in aqueous residue (mg) (%) 0. 25 4-6 0. 5 50 1. 0 70 5. 0 80 - 85

Calcium oxalate precipitate was dissolved in hydrochloric acid and re­ precipitated by addition of excess oxalic acid. This step removed phosphate satisfactorily. The aqueous residue did not show the presence of phosphate when tested with ammonium molybdate reagent. A rigorous check on removal efficiency was made by carrying out calcium oxalate precipitation in urine containing additional 1 g P as phosphate. The phosphate content in aqueous residue from TTA extraction was about 0. 2 mg. The overall recovery of 75% reported in this paper for the method is not.a maximum, and efficiencies of 85-90% have been easily achieved with careful manipulations. For routine procedure described in the Appendix, 75% recovery is the average figure obtained in the laboratory. In the time- efficiency balance this is adequate when the laboratory has to handle a large number of samples at a tim e. As a. check on routine operations as well as individual manipulative skill, the samples are often spiked with Th234 as it does not interfere with spectro- photometric observations. Th234 is a beta-emitter, and the activity in the thorium-thoronol complex solution can be measured after the spectrophoto- metric readings are taken. This control is recommended until an efficiency factor is established for the working conditions. It is useful to have in the laboratory an ion-exchange column containing natural uranium for milking Th234 for tracer studies [12].

R adium -228

Radium-228 is a low beta-emitter (0. 012 MeV) and its determination is done by counting its hard beta (1. 2 MeV) emitting daughter actinium-228. In the first instance, radium is precipitated as sulphate in a urine sample to which Pb-, Ba-carriers are added [13]. Radium is allowed to decay for two to three days to allow for the growth of daughters. Radium isotopes originating from thorium parent are Ra228 and Ra224 but chemically separated radium often contains radium from uranium series also. Of these Ra228 alone is a beta-emitter decaying to actinium. The beta-activity of the radium sulphate precipitate allowed for equilibrium actinium growth is mainly due to Ac228 and beta-emitting daughters of Ra224 (Ti 3.6 d). Radium-226 contributes betas to about 50% of its alpha-activity [14]. Table III gives growth of beta activity in 48 h after chemical separation of Ra228, Ra224, Ra22B and Ra223 [15]. PROCEDURES FOR BIO-ASSAY 203

TABLE III

BETA-ACTIVITY GROWTH {цс) IN 1000 цс RADIUM ISOTOPES AFTER 48 h OF SEPARATION

Total Radium A c Pb Bi‘ T1 daughter isotope b eta- activ ity

Ra223 - Pb211 8 8 9 .0 - 8 9 0 .4 2 6 6 9 .8

Pb207 8 9 0 .4

Ra226 - Pb214 2 9 9 . 9 Bi214 2 9 7 .4 - 5 9 7 .3 .

Ra224 - Pb212 7 2 8 .1 Bi212 4 6 4 .5 2 6 3 .6 1 4 5 6 .2

Ra228 9 9 5 .3 Pb212 0 .1 3 Bi212 0 .0 8 0 .0 5 9 9 5 .6

Measurement of Ra228 by direct counting is possible [16] if thorium-free radium in solution is set aside for the decay of Ra224 (30 d) and then taken up for radium precipitation, assuming, of course, that contributions from Ra226 are negligible. PETROV and ALLEN [IV] have separated Ac228 by solvent extraction free of other decay products. NAS - NS 3020 [18] gives several methods for Ac228 separation using solvent extraction and ion-exchange techniques. Figure 3 gives the analytical sketch for the method presented in the paper.

16-h urine + 200 mg Pb/1 5 mg Ba and +1:1 H2S04

i Precipitate , Supernatant

Dissolve in am m oniacal ED TA, . D iscard add acetic acid to pH 4. 5, centrifuge.

1 i Precipitate ' Supernatant Redissolve and precipitate D iscard as above, centrifuge.

1 1 Precipitate Supernatant

+ 5 mg La, 5 mg Pb, 5 mg Bi D iscard dry. Keep for 60 h. Dissolve in HC104, add dilute H 2S04. Allow precipitate to settle. Centrifuge.

1 Precipitate Supernatant

D issolve in am moniacal ED TA, + concentrated HC1, HF, centrifuge add acetic acid and centrifuge

, , i1 . — 1 1 1 i Filtrate Precipitate Filtrate Precipitate D iscard ' W ash and Discard Wash with HF-HC1, weigh for HF-HNO 3 , and . ’ Ba recovery water. Dry, weigh LaF3, count beta activity. Follow decay of Ac228 F ig .3

Analytical sketch for radium- 228 204 P. R. KAMATH et al.

The new method takes advantage of solubility of (BaPb)SC >4 in hot p e r­ chloric acid to release in-grown actinium in solution. Along with Ac228, radioisotopes of Pb, Bi and T1 are also released. The precipitate obtained from urine is first purified by an ammoniacal ethylene-diamine-tetraacetic acid (EDTA) treatment [13] and allowed for Ac228 growth. BaSC>4 is then dissolved in the presence of La-carrier (for Ac), and hold-back carriers Pb and Bi. Thallium-208 is short-lived (3.1 min) and does not interfere. Barium and lead sulphates are first precipitated from perchloric acid solution by addition of dilute sulphuric acid and then La(Ac)F3 is precipitated in the supernatant in presence of hydrochloric acid. The chemical recovery for La was 85-90% and separated Ac228 decayed in accordance with its half-life. The recovery of Ba was 80-90%. The overall efficiency calculated from chemical recovery as well as by counting was about 80%. Exclusive of the delay allowed for actinium growth, other operations can be completed in one day.

Radiocobalt

Most procedures for determination of Co60 utilize its gamma-activity for estimation with a kicksorter [19]. For very low concentrations as well as in the presence of other activities this method can be unsuitable or in­ convenient. Radiochemical separations of cobalt are not very satisfactory for low-level work and involve bulky precipitations using organic reagents like alpha nitroso beta naphthol or nitroso-R-salt and conversion to oxide. Alternatively, cobalt is precipitated as potassium cobaltinitrite or the cobalt activity is counted after a final electrodeposition. The method developed in this laboratory is outlined in Fig. 4. The method consists in leaching out cobalt with dilute acid from urine ash and separating it from contamination with ion-exchange column. Cobalt is pre­ cipitated as C 0 C 2 O4 • 2 H 2O dried at 100°C and counted fo r b eta -a ctivity. KRAUS and NELSON [20] have described absorption characteristics of elements from hydrochloric acid solutions on strongly basic ion exchangers. Cobalt is retained on anion-exchange column from 8 - 9N HC1 and eluted with 4N HC1. Zirconium alone interferes in this separation. Nickel is not absorbed on the column and passes through in the effluent. Zirconium phosphate is insoluble in dilute acids. Trace zirconium present in urine is eliminated by the leaching of cobalt from urine ash with dilute acids in the presence of Zr-carrier and phosphate. A decontamination factor of 104 or more was realized when the ash containing phosphates was leached with very dilute hydrochloric acid. Cobalt was quantitatively recovered. • IRA 400 (Cl*) was preferred to Dowex 1 for column preparation as the former was heavier and did not float in the medium. A 5-g resin (dry) column was set up in a burette with glass-wool plugs to hold the column. Cobalt form s an anionic complex in 8 - 9N HC1 and is absorbed on the column. It is quantitatively recovered when eluted with 50 ml 4N HC1. Cobalt oxalate dihydrate is stable up to 130°C and decomposes in stages beyond it [21]. Overall chemical recovery is 80% and the estimation can be completed in a day or so. Figure 5 is a self-absorption curve for the oxalate. PROCEDURES FOR BIO- ASSAY 205

25 ml urine - add 5 mg Zr, 10 mg Co, 50 m g P 0 4 carriers, wet ash with HNO3 and evaporate with concentrated HC1 extract with 0. Щ HC1, centrifuge.

1 1 Supernatant Residue

Evaporate to dryness, dissolve in D iscard 8 Ы HC1, pass through IRA 400 anion-exchange column. Pass further 8 £J HC1.

• Resin Effluent

Elute the column with 4N HC1. D iscard Evaporate to dryness. Extractwith 0. 2N HC1. Add Na2S0 3 , dissolve sulphite in 2N HC1. Add oxalic acid, warm, keep for \ h, centrifuge. 1 Precipitate Supernatant

Wash with water and alcohol, dry Discard at 90-100°C. Weigh (CoC 204 • 2H20 ), count beta activity

F i g .4

Analytical sketch for radiocobalt

F ig . 5

Self-absorption curve for Сои in cobalt oxalate

Radioiodine

The determination of radioiodine in urine has been carried out by an initial oxidation of the organic matter, solvent extraction of elemental iodine and subsequent precipitation as silver iodide [22], FLETCHER [23] fractio­ nated urinary iodine into inorganic iodide and organic-bound iodine. Eighty to ninety per cent of the total has been reported as inorganic iodine in his 206 ■P. R. KAMATH et a l.

500 ml urine + 10 mg I carrier + 2N H N 03, 200 mg A gC l, stir for 30 min, centrifuge

1 Residue Supernatant Dissolve in ammonia, centrifuge D iscard and wash , 1 1 Residue Supernatant

D issolve in 2% KCN, + 10 mg Z r D iscard carrier, precipitate Zr (OH )4 with ammonia, centrifuge

1 Supernatant Precipitate

Add drop by drop 2£í HNO 3 to D iscard precipitate Agi, wash free of acid, dissolve in KCN and reprecipitate - wash, dry and weigh. Count beta activity.

F ig . 6 .

Analytical sketch for radioiodine studies. It has been also claimed that it was possible to demonstrate that nearly 95% of the urinary iodine was inorganic in internal exposures of the inorganic iodine [24] . The method outlined in Fig. 6 gives a rapid deter­ mination of iodide in urine and is based on a method [25] developed in the laboratory for determination of iodine in sea water. On account of the high insolubility of silver iodide in water compared with silver chloride (solubility product is 106 times lower), iodide in solution quickly replaces chloride from silver chloride used as absorbent. For the 10-mg iodide carrier used in the method the exchange was completed in 20-30 min and the efficiency was not influenced by increasing the amount of the absorbent. The recovery was same for silver chloride content of 200-1000 mg. The chemical as well as the radiochemical recoveries were about 80% for iodide in solution. An experiment was carried out spiking urine with solutions containing mixed fission products as well as radioiodine. The gamma spectrum showed only radioiodine. When tested with zirconium activity in the presence of Zr-carrier it was found that zirconium is retained only partially and by physical adsorption on the absorbent. Zirconium is eliminated when silver iodide is cleaned by cyanide complexing and reprecipitation. The method is rapid and can be completed in three to four hours and gives a measure of inorganic radio­ iodine present in urine. •

Radiocaesium , ■

Very elaborate separation schemes for Cs1з^ have been reported [26] utilizing adsorbent columns followed by carrier precipitations and mounting of Cs as chloroplatinate finally for beta counting. Other schemes utilize separation as an alum, chlorostannate, perchlorate, phosphomolybdate and molybdoarsenates. Cs2PtC l6is popular for final separation and counting as it gives very high decontamination from other fission products [27]. PROCEDURES FOR BIO-ASSAY 207

100 ml urine + 50 mg P0 4 and 10 mg Cs. Raise the pH to 8 , boil, centrifuge ______!_____ ‘ ' I I Supernatant Precipitate •

+ HNO3 , II2O2 and boil. Cool and Discard - dilute (pH 1), add 500 mg A M P , stir for 30 min, centrifuge. .

Precipitate Supernatant

Dissolve in minimum ammonia, boil, Discard dilute to about IN in ammonia, pass through ammoniacal Dowex 50, wash with water. '

- Resin Effluent and

Elute with 6 N HC1 and evaporate washings the eluate. Heat with HNO3 and Discard finally flame off ammonium salts. ' i Residue ' .

Dissolve in 6 N HCl and precipitate CSgPtCly. Wash with 6 N HCl, twice with hot water, transfer with ’ . alcohol, dry, weigh and count beta activity ' «g -7 . .

Analytical' sketch for radiocaesium

The analytical procedure developed in this laboratory is outlined in F ig . 7. ■ Ammonium phosphomolybdate (AMP) is stirred in urine containing Cs-carrier. Caesium (and Rb) is absorbed by exchange from solution and not potassium. Zirconium and ruthenium are reported to be absorbed in the process [27] . Rubidium contribution from fission is small (1. 6X 10‘4% yie ld ) and in estim ation of Csi37 there would be little in terferen ce from Rb. Rubidium interference could be significant when urines are examined for induced Cs134 activity. Caesium chloroplatinate is 11 times more insoluble than rubidium salt and 4 times that of potassium chloroplatinate [28] . An initial precipitation of urinary calcium as phosphate removes zirconium. The supernatant is heated with concentrated nitric acid and peroxide to oxidize organic matter. In the absence of peroxide oxidation AMP turns green. AMP is dissolved in ammonia and the solution is passed through an ammoniacal Dowex-50 column. In this process ruthenium might be expected to be eliminated in a complexed form as aminocomplexes of Ru are w ell known. , The recovery for the method is 80% and the procedure gives a clean separation from Zr and other fission products. 500 mg AMP are necessary for recovery of added 10 mg Cs-carrier. Studies are still in progress for separation of Rb and Cs using ion- exchange resin and HCl-solvent elutions.

RE SU LTS ■

Experiments were carried oht with spiked urine samples to determine the actual recoveries for the methods presented in the paper. The results 208 P. R. KAMATH e t a l.

TABLE IV

RECOVERY STUDIES WITH SPIKED URINE

Experimental Spiked activity Actual recovery Recovery procedure (cp m ) (cp m )

3 3 . 0 * 2 7 .5 8 3 .4

1 1 2 . 0 . 8 3 . 5 7 4 .6

Thorium (nat) . 1 3 8 .0 8 7 .2 ' 6 3 .2

3 1 8 .5 2 8 0 .0 8 8 . 0

3 3 5 .0 2 3 9 .5 7 1 .4

1 1 .3 1 0 . 0 8 8 .5

2 2 . 6 1 8 .8 8 3 .2

Ra228 2 2 . 6 1 7 .4 7 7 .0

2 8 .0 2 0 . 6 7 3 .6

' 5 6 .5 5 0 .0 8 8 .5

2 2 8 .0 1 8 9 .0 8 3 .0

2 2 3 .0 1 7 3 .0 7 7 .6

C o 60 • 2 2 6 .0 1 7 9 .0 7 9 .2

2 2 6 .0 1 5 0 .0 6 6 .5

2 2 7 .0 1 8 1 .0 7 У. 7

5 0 .5 4 1 .5 8 2 .2

7 8 .5 6 4 .8 3 2 .5

[131 9 2 .0 6 7 .5 7 3 .4

1 9 6 .0 1 5 8 .0 8 0 .6

7 8 5 .0 6 5 5 .6 8 3 .5

1 2 6 .0 1 0 1 . 0 8 0 .2

1 8 0 .0 1 2 6 .0 7 0 .0

C s 137 6 3 1 .0 4 6 7 .0 7 4 .0

- 6 3 1 .0 4 5 6 .0 7 2 .3

6 5 5 .0 4 8 2 .0 7 3 .6

* T h 234

are given in Table IV. The values reported are observed recoveries, un- corre'cted for chemicàl efficiency of the experiment.' PROCEDURES FOR BIO-ASSAY 209

ACKNOWLEDGEMENTS •

We are grateful to Dr. A. K. Ganguly, Head, Health Physics Division, for guidance and interest in these studies and to Shri A. S. Rao, Director of our Group for constant encouragement. We wish to thank the IAEA for facilities given for oral presentation of the paper at the Symposium.

APPENDIX

I. DETERMINATION OF THORIUM (nat)

Reagents

Oxalic acid solution: Saturated and 3% in water Oxalic acid wash solution: 3% in 0 . IN HC1 Nitric acid solution: Concentrated; 2N; IN and 0.1N Perchloric acid: Concentrated (60%)

Perchloric acid solution: pH 0.8 - 1.2 (15 ml of 60% HC104 /1 ) Hydrochloric acid: Concentrated; 4N TTA solution in benzene: 10 g of TTA dissolved in 100 ml of pure benzene and stored in brown bottle . Thoronol reagent: 25 mg of thoronol dissolved in 100 ml water. The solution is freshly prepared every week Standard thorium solution: A solution of pure thorium nitrate is prepared and standardized gravimetrically. Solution

containing 1 jig of thorium per ml is prepared •. by diluting the solution. Ca-carrier solution: 100 mg Ca/ml

Procedure

(1) Measure the volume of the urine sample and transfer the sample to a 2-1 Pyrex beaker. Add(a) 3 ml Ca-carrier and adjust(b) the pH to 3. Heat the solution to boiling. (2) Add 15 ml saturated‘oxalic acid solution to every 500 ml of urine gradually and with stirring; boil for further 5 min. Allow to cool and settle (3-4 h). . (3) Siphon out most of thè supernatant to a separate beaker(c) till not more than 5 ml is left with the preci­ pitate (d). (4) Add 10 ml of concentrated HC1 to the beaker and dissolve the oxalate precipitate by heating. Cover with a watch glass and heat to boiling. ' , (5) Add to the solution 300 ml of hot 3% oxalic acid solution, stir, boil for 2 min, cool and allow the pre­ cipitate to settle overnight.

( 6 ) Decant the supernatant and transfer the precipitate to a 40-ml centrifuge tube with oxalic acid wash solution. Centrifuge at about 2000 rpm for 5 min. (7) Transfer the supernatant and wash the precipitate first with 10 ml oxalic acid wash solution and then' with water.

( 8 ) Transfer the precipitate quantitatively along with nitric acid washings of the beaker into a 25-ml silica

dish. Add 5 ml perchloric acid and heat till no more HC104 fumes are evolved. Remove the last traces of perchloric acid by heating on a wire gauze for a few minutes. (9) Dissolve the residue in 2 ml IN nitric acid, and transfer the solution to a 100-ml-separating funnel con­

taining 20 ml TTA solution. Wash the dish three times with . 6 ml water and transfer the washingsfe) to the separating funnel. Shake mechanically for 15 min. Allow to settle for 15 min.

(10) ‘ Withdraw the aqueous layer(c); wash the organic layer with 5 ml 0.1N HN03 followed by 10 ml water. Shake for 5 min, and allow about 5 min for separation of layers. Separate the aqueous layer.

14 2 10 P. R. KAMATH et a l.

(11) Add 20 ml 2N HN03 to the TTA solution in the separating funnel, shake for 15 min and allow to settle for 15 min. Collect the aqueous layer in a 50-ml silica crucible. Wash the sides of the funnel with 5 ml

water. Collect the washings in the silica crucible. Add 1 ml HC104 to the crucible and evaporate till no more fumes of perchloric acid appear. Cool. (12) Evaporate 1 ml of standard thorium solution similarly with 1 ml perchloric acid in a separate crucible. (13) Add 4 ml dilute perchloric acid (pH 0.8 - 1.2) to the crucibles to dissolve the residue. Add 1 ml thoronol reagent. Stir and allow to stand for 10.min. Prepare a reagent blank similarly by adding 1 ml thoronol reagent to 4 ml perchloric acid. ' (14) Transfer the solutions^) to Beckman cells of 1 cm light path. Read the absorbency against the reagent blank at 545 nm in the Beckman spectrophotometer.

Calculations

= absorbency of the standard .

a 2 = absorbency of the sample

c [ = amount of thorium standard taken

. ‘ c 2 ~ amount of thorium in the sample .

.2l - " c 2 a2 '

Notes .

(a) It is useful to add Th 234 activity to determine the efficiency of the method. On an average 7 5 - 8 0 *7 0 recovery has been observed. (b) pH adjustment for fresh sample is done by adding HC1. If the sample is acid preserved, addition of alk ali (2NJ NaOH) is necessary. (c) If an estimation of Ra is to be made in the sample, the supernates and washings from Ca oxalate sepa­ rations should be preserved. Aqueous residues from Th - TTA extraction also contain about 10% of total Ra present. . (d) . In siphoning, if the precipitate is disturbed, it may be allowed to settle or separate by centrifuging. Normally this can be avoided by careful manipulation. (e) Maximum recovery in single extraction has been observed when there are equal volumes of organic and aqueous layers. The aqueous volume should be adjusted to 20 ml.

(f) If the urine is spiked with Th 234 the solution should be preserved for beta-activity determination. For this the solution is evaporated to dryness with HClp 4 and Th is precipitated as fluoride with a La-carrier.

LaF3 is taken up for beta-counting.

II. DETERMINATION OF RADIUM-228

R eagents

La-carrier solution: 10 mg La/ml Ba-carrier solution: 10 mg Ba/ml Pb-carrier solution: 100 mg Pb/ml Bi-carrier solution: 100 mg Bi/ml . Perchloric acid: . Concentrated (60%) ' Hydrofluoric acid: Concentrated (40%) HF - HC1 wash solution: 14.8 ml concentrated HF, 34.0 ml concen­ trated HC1 - diluted to 400 ml with distilled water PROCEDURES FOR BIO-ASSAY 211

HF - H N 0 3 wash solution: 14.8 ml concentrated HF, . 25.2 ml concentrated HNQj diluted to 400 ml with distilled water Hydrochloric acid: Concentrated (X0N) . Sulphuric acid: 4N, m EDTA solution: 0.25 M ( di- sodium salt)

P rocedure

(1) Transfer the 16-h urine sampled) to a 2-1 Pyrex beaker and add 200 mg Pb-carrier per litre of urine

and 5 mg Ba-carrier. Add 20 ml of 1 : 1 H?S0 4 and stir; allow the precipitate to settle. (2) Decant the supernatant and transfer the precipitate to a centrifuge tube and wash it with IN H,S04, and then with water. (3) Dissolve the precipitate in ammoniacal EDTA solution (5 to 10 ml), heating on a water bath if necessary. Pxecipitate BaS0 4 with the addition of acetic acid till the pH of the solution is about 4 .5 . Centrifuge and

separate the precipitate. Redissolve BaS0 4 in ammoniacal EDTA and precipitate with the addition of acetic

acid. Centrifuged3), wash the precipitate with water. Add to the BaS0 4 in the.centrifuge tube 5 mg o f Pb- and 5 mg of Bi-carriers. Slurry the contents in the centrifuge tube and evaporate to dryness under an infra-red lamp. Set aside for 30 h(c) or preferably for 60 h. (4) At the end of 30 h add 5 ml perchloric acid to the tube and dissolve the residue by gentle heating.' Digest for 5 min and then add 4 ml of diluted sulphuric acid (4N), (Note time, tn). Allow the precipitate to settle. Centrifuge and wash the precipitate with 4N HjSCJ and then with water. Collect supernates and washings in a lusteroid tube. (Set aside the precipitate of Ba, Pb sulphates(d). (5) • Add 10 ml concentrated HCl and 5 ml HF, stir and keep aside for 5 min, centrifuge and wash the preci­

pitate (LaF3) twice with HF - HCl (5 ml) solution, once with 5 ml HF - HN03 solution and twice with 10 ml of distilled water.

( 6 ) Transfer the LaF3 precipitate slurrying with water on to a tared stainless-steel pianchet, dry under infra-red lamp and count beta activity and weigh. Note the time lapse from precipitation time (step 4) to mid point of counting (t). Check for decay of Ac228.

Calculations

с = observed counts/min above background t = time lapse in hours from precipitation (step 4) to mid point of counting

E = efficiency °¡o of the counting set-up for Ac 228 R = chemical recovery of La% . S = chemical recovery of Ba% ‘ . F = equilibrium with parent (about 90%) in 30 h % '

с x 1 0 0 x 1 0 0 x 1 0 0 Ra228 (t 0 ) = ------;------fiC (16-h sample) E xR xSxFx2.2xl0sxexp (--^jp )

N otes .

(a) The sample as such or supemates from thorium method may be used. The PbClj may precipitate out

when Pb-carrier is added and in that case it can be dissolved by adding HNO3 ,

(b) If organic matter is present, it can be ashed with HN0 3 - H C 10 4 and then BaS0 4 precipitated com­ pletely by using HgSO*.

( c ) 30 h are allowed for Ac228 to grow into Ra228 to 90% equilibrium. At 60 h, i.e ., 10 half-lives .equilibrium will be almost complete.

(d) The precipitate contains PbS04, BaS0 4 and co-precipitated radium. BaS0 4 is separated from it and checked for Ba-recovery. Most experiments showed that this check gives 90%. recovery. The precipitate can be used to determine Ra224 as well as Ra2 2 6 and to milk Ac228 again for a repeat determination. 212 P.R. KAMATH et al.

III. DETERMINATION OF RADIOCOBALT

R eagents

Anion-exchange resin: IRA 400 (С Г); 25-50 mesh-5 g (dry) resin is supported over glass wool in a burette and

washed with 8 N HC1 ' . Oxalic acid: Saturated solution

Hydrochloric acid: 0 .1 N , 4 N , 8 N and c o n ce n tra te d Nitric acid: Concentrated Sodium sulphite: A.R. C o - c a rrie r: 10 mg Co/ml Zr-carrier: 10 mg Zr/ ml Potassium hydrogen phosphate solution: 50 mg P04/m l

Procedure

(1) Take 25 ml of a 16-h urine sample in a silica dish. Add 1 ml of Co- and 1 ml of Zr-carrier, and

evaporate the solution to dryness. Add 5 ml of concentrated HN03 and wet ash the residue. Add 3 ml of concentrated HC1 and 2 ml phosphate solution, stir and evaporate to dryness. (2) Add 5 ml 0.1N HC1 solution to the residue, digest and transfer the solution into a 15-m l centrifuge tube; re-extract with 5 ml 0.1N HC1, wash and add to the centrifuge tube and centrifuge. Transfer supernatant to a 50-ml beaker. Wash twice with 0. IN.HC1. Evaporate supernatant and washings to dryness. Cool and

dissolve the residue in 3 ml 8 N H C l(a ). (3) Set up an anion-exchange column in a burette, with 5 g of dry IRA-400 resin. Wash the resin with

hydrochloric acid and condition it with 8 N HC1.

(4) Pass the 8 £J acid solution of the residue through the column at 0. 5 ml/min. Pass a further three 1 .5 -ml

and one 4 -ml 8 N HC1 washings of the beaker followed by one of 25 ml 8 N HC1. (5) Elute the column with 50 ml 4N[HC1 at 1 ml/min(b). Collect the eluate In a 100-ml beaker.

( 6 ) Evaporate the eluate to almost dryness. Dissolve the residue in 1 ml 0.2N HC1 and transfer into a 40-ml centrifuge tube, wash the beaker with 10 ml water and add to the tube. Add 300-400 mg sodium sulphite and stir; heat on a water bath till cobalt sulphite(c) precipitates. (7) Add drop by drop 2N HC1 to dissolve sulphite precipitate and then 5 ml saturated oxalic acid solution. Keep for half an hour on a boiling water bath with frequent stirring(d). Cool, centrifuge.

( 8 ) Discard the supernate; wash the precipitate once with water. (9) Transfer the precipitate to a previously.washed, dried and weighed Whatman-542 filter paper(2.5 cm2) on a demountable filter assembly. Wash with alcohol, ether and dry at 80 - 100°C in an oven, weigh(e), and count beta activity. .

Calculations

V = total sample volume of 16-h void Y = chemical yield of Co °}o . С = counts/min above background(^) E = efficiency of counter %

C x V x 100 x 100 Coso = ------цС

- л , , 2 5 x Y X E X 2 . 2 х 1 0 6 (16-h sample) . •

N otes ‘

(a) Urinary NaCl dissolves with difficulty in 8 N HC1. In the following acid washings cobalt is completely leached from the chloride residue. (b) The resin column can be re-used after regeneration by passing dilute HC1 and finally washing with water. PROCEDURES FOR BIO-ASSAY 213

(c) At this stage cobalt sulphite separates and the solution turns turbid. (d) The precipitate forms slowly. Frequent stirring is helpful.

(e) The oxalate precipitate has the formula - CoC 20 4' . 2 H20 ; it dehydrates above 130°C. (f) The background correction should include chemical background also. . (g) Self-absorption correction is made using a self-absorption and scattering curve for cobalt oxalate. (Fig.5).

IV. DETERM INATION’OF RADIOIODINE:

Reagents .

Iodide carrier: 10 mg I/ml . Zr-carrier; 10 mg Zr/ml Silver chloride: Precipitated in nitric acid medium, washed, dried and ground to 50 mesh

Ammonia solution: 0 . 8 8 liquor ammonia

Potassium cyanide: 2 °jo aqueous solution Nitric acid: 2N

Procedure ’

(1) To 500 ml urine add 10 mg iodide carrier, 10 ml 2N nitric acid and 200 mg AgCl. Stir for 30 min

and let it settle for 2 h. (2) Decant the supernatant and centrifuge the residue. Wash with 2N nitric acid and then with distilled w ater. (3) Add 10 ml ammonia.solution and dissolve the silver chloride. Centrifuge and separate the silver iodide. Wash the iodide with 5 ml of ammonia and then with water. • (4) Dissolve the Agi in 5 ml KCN solution and add 0.5 ml Zr-carrier. Add drop by drop a-mmonia and

stir, till Zr(OH )4 precipitate alone remains^).

(5) Separate Zr(OH )4 by centrifugation. To the supernate in a separate centrifuge tube add drop by drop 2N nitric acid till the Agi is precipitated. Centrifuge and wash the precipitate with 2N nitric acid and water.

( 6 ) Redissolve the precipitate in 5 ml KCN solution and centrifuge off any residue present. Precipitate the Agi again in the supernatant with 2N nitric acid. Wash the Agi twice with nitric acid and then with water. (7) Transfer the residue on a tared filter paper (2.5 crrf), washed and dried previously, using demountable filter assembly. Wash with alcohol, ether and dry at 100°C in an oven for 5 min and weigh. Count beta activity. Note counting time to calculate decay factor.

Calculations

С = observed beta-activity counts/min above background Y = chemical recovery of iodine carrier °]o t = time in days between voiding of urine to counting the precipitate V = total urine voided

E = efficiency of the instrument for I 131 beta activity °Jo

I1 3 1 = ______С x 100 x V x 100 ______fdC

(16-h sample) E x 500 x Y x {exp (- )} x 2.2 x 10 6

N ote

(a) Organic matter which has been carried along till this stage is eliminated along with Zr(OH )4 precipitate. 214 P.R. KAMATH et al.

V. DETERMINATION OF RADIOCAESIUM

Reagents

Cation-exchange resin: 8 g of Dowex 50, 50- 100 mesh in a burette supported over glass wool. Ammoniacal form o f the resin is prepared by passing 100 ml of IN ammonia “ Nitric acid: Concentrated • Hydrochloric acid: 6N .Ammonia: 3N, IN Ammonium Phospho Molybdate: •A.R, (AMP) Chloroplatinic acid: 5% solution in water Hydrogen peroxide: 30% vol./vol. Cs-carrier: ' 10 mg Cs/ml Potassium-hydrogen phosphate solution: 50 mg P04/m l

Procedure

(1) Take 100 ml of urine.from a 16-h sample in a 400-ml beaker. Add 10 mg Cs-carrier (1 ml) and 50 mg of phosphate solution and raise the pH to precipitate calcium phosphate, allow to settle. Separate by centri­ fuging. Wash, take the supernate and washings in another beaker. Add 10 ml of concentrated HN0 4 and 10 ml hydrogel peroxide. Heat gradually to boiling. Cool to room temperature. (2) Add 500 mg AMP and stir for 30 min and allow to settle. (3). Decant the supernatant. Centrifuge the residue in a 40-ml tube. Wash the AMP with IN nitric acid and then with water.' (4) Add 10 ml 3N ammonia and heat to solution. Dilute the solution with 20 ml water and centrifuge if the solution is cloudy. (5) Pass the supernatant solution (about IN in ammonia) through ammoniacal Dowex-50 column conditioned in IN ammonia solution, at the rate of 0.5 ml/min. Pass a further 15 ml IN ammonia through the column.

( 6 ) Wash the column with about 100 ml distilled water at 1 ml/min until the effluent is no more ammoniacal.

(7) . Elute Cs from the column with 50 ml 6 N HCl.

( 8 ) Collect the eluate in a beaker and evaporate to dryness, with 5 ml of concentrated HNO3 . R em ove any remaining ammonium salts by directing a small ñame on the sides of beaker.

(9) Dissolve the residue in 2 ml 6 N HCl and transfer completely with 6 N HCl (4 ml) to a 15-ml centrifuge tube, warm on hot water bath. Add 0.5 ml 5% H2 PtCl6, stir well, and allow to settle for 5 min. Centrifuge

and wash the chloroplatinate with 5 ml 6 N HCl, 1 ml water and twice with alcohol. (10) Transfer the precipitate with alcohol to a tared Whatman-542 filter paper(2.5 cm2), washed and dried previously, on a demountable filter assembly. Wash with ether and dry in an oven for 5 min. Weigh, and count beta activity(a).

Calculations

С = counts/min observed beta activity background E = efficiency for counting set-up % V = total void for 16-h sample Y = chemical recovery of Cs-carrier %

С X 1 0 0 x V x 100 • CS137 = E X 100 X Y x 2.2 x 10 6 Ц° ( 1 6 - h sam ple)

Note

(a) If high activities are observed it is useful to obtain a gamma spectrum to distinguish Cs137, Cs 134 and any contamination from Rb-activity. PROCEDURES FOR BIO-ASSAY 215

REFERENCES -

[1] BEIERWALTER, W.H. , JOHNSON, P.C. and SOLARI, A .J., Clinical uses of radioisotopes, W.B. SAUNDERS & C o ., London (1957), (See under different radioisotopes). [2] MIDDLESWORTH, L .V ., "Factors influencing thyroid uptake of iodine isotopes from nuclear fission - A review", Hlth Phys. 9 12 (1963) 1200. . [3] BOECKER, B .B ., Thorium inhalation studies, UR-605 (1961-62). [4] REYNOLDS, J.C ., GUSTAFSON, P.F. and MARINELLI, L.D ., Retention and elimination of radium isotopes produced by the decay of thorium parents within the body, ANL - 5689 (1967). [5] SUNTA, C.M . and NAMBI, К .S., Safe limits of urinary excretion of some radionuclides, Indian Science Congress (1963). ‘

[ 6 ] MEHTA, S.K ., MAHAJAN, N.M. and KAMATH, P.R., S-35 accident in isotope division, АЕЕТ/НР/ Survey/34. (1960). [7] EISENBUD, M. et a l., Naturally occurring radionuclides in foods and waters from the Brazilian areas of high radioactivity, International-Symposium on 'The Natural Radiation Environment', Rice University, Texas (1963), Chicago Press N .Y ., to be released.

[ 8 ] KAMATH, P.R. and SOMAN, S.D ., Bioassay procedures at Trombay Establishment, AEET/HP/BA/l ( 1 9 5 8 ). [9] International Commission on Radiological Protection, Report of Committee II (1959).

[10] SCHUBERT, J., "Estimating radioelements in exposed individuals”, Nucleonics 8 3 (1958) 76. [11] BROOKS, R .O .R ., "Biological monitoring of persons working with radioelements", Brit. J. clin. Practice

14 6 (1960) 465-473. . [12] BERMAN, S.S. e ta l., Separation of carrier free thorium-234 from uranium by anion exchange, Talanta 4 (1960) 153-157. . [13] GOLDIN, A.S.., "Determination of dissolved radium", Anal. Chem. 33 3 (1961) 406-409. [14] GOPINATH, D.V. and HARI Singh, Activity build up of naturally occurring radioactive materials; Part 1, AEET/HP/Th/3 (1960). , [ 1 5 ] HARI Singh, com m unication. [16] KAMATH, P.R. et a l., Environmental natural radioactivity measurements at Trombay Establishment, International Symposium on 'The Natural Radiation Environment' , Rice University, Texas (1963), Chicago Press N .Y ., to be released. [17] PETROV, H. and ALLEN, R ., Anal. Chemv 33 (1961) 1303. [18] NAS-NS 3020, Radiochemistry of rare earths, scandium, yttrium and actinium, NRC-Publication, USAEC (1961). ' [19] NAS-NS 3041, Radiochemistry of cobalt, NRC-Publication, USAEC (1961). [20] KRAUS, K.A . and NELSON, F ., "Anion exchange studies of fission products" Proc. Int. Conf. PUAE 7 (1956) 118. . ' [21] GARN, P.D. and KESSLER, J.E ., “Thermogravimetry in self generated atmospheres" Anal. Chem. 32 (1960) 1565. ‘

[22] MARRIOTT, J .E ., "The determination of I«i in urine1,1 Analyst 84 (1959) 33-37. . [23] FLETCHER, K .Fractionation of urinary iodine”, Biochem. J. 67 (1957) 136. [24] JOYET, G ., "The dynamics of radioactive iodine in.normal and pathological thyroid function" Proc. Int. Conf. PUAE 10 (1956) 283. •[25] SABU, D .D ., Iodine-131 in sea water, Indian Science Congress (1963).

[26] MORGAN, A ., "Separation of caesium in sea water", Hlth Phys. ,9 8 (1963) 659. [27] EWING, A ., Determination of cesium with chioroplatinic acid in redox solutions, HW-18146 (1950). [28] WHITNEY, I.B ., Manual of standard procedures, HASL, NYO 4700 (1959).

DISCUSSION

H. WELLMAN: How do you prevent gaseous iodine formation in your procedure? P.R. KAMATH: We don't boil the urine sample for iodine at all. The urine sample is taken and we add silver chloride and collect iodine as silver iodide.

COMPARISON OF EXCRETION ANALYSIS WITH WHOLE-BODY COUNTING FOR ASSESSMENT OF INTERNAL RADIOACTIVE CONTAMINANTS

C.W . SILL, J.I. ANDERSON AND D.R. PERCIVAL HEALTH AND SAFETY DIVISION, ' UNITED STATES ATOMIC ENERGY COMMISSION, IDAHO FALLS, IDAHO, UNITED STATES OF AMERICA

Abstract — Résumé — Аннотация — Resumen

COMPARISON OF EXCRETION ANALYSIS WITH WHOLE-BODY COUNTING FOR ASSESSMENT OF INTER­ NAL RADIOACTIVE CONTAMINANTS. Since the beginning of the atomic energy era, the detection and determination pf internal contamination by radionuclides has depended primarily on the routine examination of urine. During the past three years, the in vivo whole-body counting programme being carried out at the National Reactor Testing Station has demonstrated clearly that urinalysis is grossly inadequate as a general ' monitoring technique for internai contaminants. Thirty-one different radionuclides have been encountered in human subjects by whole-body counting, of which only iodine-131-133, caesium -134-137, mercury-197-203 and molybdenum-technetium-99 were excreted significantly in the urine. The other nuclides were eliminated so exclusively in the faeces that, except for an occasional trace, the nuclide could not be detected even in a 1500-ml urine sample. The conclusion seems clear that for inhalation of particulates, the metabolic fate will be more dependent on the physical properties of the particle with which the nuclide is associated than with the chemical properties of the element itself. The effective half-lives and modes of excretion are pre­ sented for many of the nuclides encountered. In vivo whole-body counting avoids many of the problems of excretion analysis for gamma emitters. However, whole-body counters are generally expensive both in terms of the original cost of the. equipment and the continuing costs of working time lost while the subjects being counted are away from their jobs. A new instrument is described in which the heavy shielding normally associated with whole-body counters is eliminated to achieve portability and reasonable costs that most laboratories can afford. The prototype weighs about 640 pounds and can be mounted in a small van to permit moving the counter to different reactor sites to increase the coverage for the same cost. Yet, approximately 0.01 /je of caesium -137 can be detected

in a 1 0 - min count, a level that is more than adequate for practical personnel protection.

DÉTERMINATION DE LA CHARGE CORPORELLE: COMPARAISON DE L’ANALYSE DES EXCRÉTIONS ET DE L’ANTHROPOGAMMAMETRIE. Depuis l'avènement de l’ère atomique, la détection de la contamination interne et la détermination de la charge corporelle ont reposé essentiellement sur l’examen régulier des urines. Au cours des trois dernières années, l'exécution d'un programme d'anthropogammamétrie in vivo, à la Station nationale d'essai des réacteurs, a clairement démontré que l'analyse des urines était loin de constituer une méthode satisfaisante de contrôle de la contamination interne. Sur les 31 radionucléides différents qui ont été décelés chez des êtres humains par anthropogammamétrie, seuls l'iode 131-133, le césium 134-137, le mercure 197-203 et le molybdène-technétium 99 sont excrétés en quantités appréciables dans les urines. Les autres radionucléides sont exclusivement évacués dans les matières fécales et, à l'exception de traces occasion­ nelles, on n'a même pas pu en déceler dans un échantillon d'urine "de 1500 cm3. Il semble donc clairement en ressortir que le sort métabolique d’une substance particulaire inhalée dépendra davantage des propriétés physiques de la particule à laquelle est associé le nucléide que des propriétés chimiques de l'élément lui-même. Les auteurs indiquent la période réelle et les modes d’excrétion d'un grand nombre de radionucléides décelés. L*anthropogammamétrie in vivo supprime un grand nombre des problèmes inhérents au dosage des émetteurs gamma dans les excrétions. Toutefois, l'anthropogammamétrie est une méthode généralement coûteuse, tant en raison du prix de l’appareil que des heures perdues par les sujets suivis qui doivent se rendre loin du lieu de leur travail. Les auteurs décrivent un nouvel appareil dans lequel on a supprimé le lourd blindage dont un anthropogammamètre est normalement muni, de manière à le rendre portatif et assez bon marché pour la plupart des laboratoires. Le prototype pèse environ 320 kg et on peut le fixer sur une petite camionnette, ce qui permet de transporter l'appareil dans plusieurs centres de réacteurs, et ainsi d’accroître

217 218 С. W. SILL et al. le nombre des personnes examinées sans augmenter les dépenses. Actuellement, on peut déceler, après une période de comptage de 10 min, environ 0, 01 дс de césium 137, rendement plus que suffisant pour assurer en pratique la protection du personnel. .

СРАВНЕНИЕ ДАННЫХ АНАЛИЗА ВЫДЕЛЕНИИ.С ДАННЫМИ ИЗМЕРЕНИЯ РАДИО­ АКТИВНОСТИ ВСЕГО ОРГАНИЗМА С ЦЕЛЬЮ ОЦЕНКИ ВНУТРЕННЕГО РАДИОАКТИВНОГО ЗАРАЖ ЕНИЯ. С самого начала эры атомной энергии обнаружение и определение внутреннего заражения радиоизотопами зависело главным образом от обычного исследования мочи. Осу­ ществление в течение последних трех лет на Национальной станции по испытанию реакторов программы измерения радиоактивности всего организма in vivo ясно показало недостаточность этого метода как общей методики измерения внутреннего заражения. При измерении радио­ активности всего организма у людей обнаружен 31 радиоизотоп, из которых только йод-131-133, цезий-134-137, меркурий-197-203 и молибден-технеций-99 выделялись с мочой в значитель­ ном количестве. Другие изотопы выводились исключительно с калом, так что, за исключением случайных, следов, изотоп невозможно было обнаружить даже в образце мочи объемом 1500 мл. Напрашивается вывод, что при выдыхании частиц метаболический процесс будет больше за­ висеть от физических свойств частицы, с которой связан изотоп, чем от химических свойств самого элемента. Представлены эффективные периоды полувыведения и формы выделения для многих из рассматриваемых радиоизотопов. При измерении радиоактивности всего оргенизма in vivo устраняются многие проблемы анализа выделений для определения гамма-излучающих изотопов. Однако счетчики для из­ мерения радиоактивности всего организма в общем являются дорогостоящими ка с точки зре­ ния первоначальной стоимости оборудования, так и ввиду постоянной стоимости теряемого пациентами рабочего времени в период измерения у них содержания радиоизотопов в организме. Описывается новый прибор, в котором устраняется тяжелый экран, обычно имеющийся на таких счетчиках, с целью обеспечения портативности и разумной стоимости прибора, которую могут позволить себе большинство лабораторий. Прибор весит около 320 кг и может быть смонтиро­ ван в небольшом вагоне, позволяющем перевозить счетчик на разные реакторные площадки для увеличения эффекта использования при той же стоимости. Тем не менее, при 10-минутном подсчете можно обнаруживать гамма-излучающие радиоизотопы в количестве до 0 , 0 1 мккюри цезия-137; это более чем достаточный уровень для практической защиты персонала.

EVALUACIÓN DE LOS CONTAMINANTES RADIACTIVOS INTERNOS: COMPARACIÓN DE LOS RESUL­ TADOS DEL ANÁLISIS DE LAS EXCRECIONES Y DE LA ANTROPOGAMM AMETRÍA. Desde el comienzo de la era de la energía atómica, la detección y determinación de la contaminación interna provocada por los radionúclidos se ha basado principalmente en análisis periódicos de orina. En los últimos tres años, el pro­ grama de antropogammametría in vivo que se desarrolla en la National Reactor Testing Station ha demostrado claramente que el análisis de orina dista mucho de ser eficaz como técnica de vigilancia radiológica general de los contaminantes internos. Efectivamente, la antropogammametrfa ha demostrado la presencia en el cuerpo humano de 32 radionúclidos diferentes de los cuales solamente el yodo-131-133, el cesio-134-137, el mercurio-197-203 y el molibdeno-tecnecio-99 se excretan en cantidad significativa en la orina. Los núclidos restantes se eliminan de modo tan exclusivo por las heces que, salvo algún vestigio ocasional, no podría descubrirse su presencia ni aun en una muestra de orina de 1500 mi. Se ímponé, pues, la conclusión de que, en lo que respecta a la inhalación de macropartículas, el curso de metabolismo dependerá más de las propiedades físicas de la partícula con que está relacionado el núclido que de las propiedades químicas del elemento en sí. Los autores presentan los períodos efectivos y los modos de excreción para muchos de los núclidos hallados. Con la antropogammametría, se evita gran parte de los problemas que plantea el análisis de las ex­ creciones en la determinación de emisores gamma. Sin embargo, los antropogammámetros son generalmente onerosos, tanto si se considera el precio original del equipo como los costos correspondientos al tiempo de trabajo perdido mientras los sujetos sometidos al recuento permanecen lejos de sus puestos. Los autores describen un nuevo instrumento en el cual, a fin de facilitar su transporte y reducir el precio a lo que puede pagar razonablemente la mayoría de los laboratorios, se ha eliminado el pesado blindaje que normalmente requieren los antropogammámetros. El prototipo de este nuevo aparato pesa aproximadamente 640 libras y puede montarse sobre un pequeño vehículo que permite transportarlo a diferentes lugares en el reactor para aumentar las apli­ caciones sin incrementar el costo. El aparato permite, no obstante, detectar con un recuento de 10 min 0,01 дс de cesio-137, esto es, un valor más que suficiente para asegurar prácticamente la protección del p erson al. COMPARISON EXCRETION ANALYSIS WITH WHOLE-BODY COUNTING 219

1. INTRODUCTION

Since the beginning of the atomic energy era, detection and determination of internal contamination from radioactive nuclides has been accomplished almost universally by examination of urine. Although the possibility of in­ halation or ingestion of insoluble particulates with subsequent elimination through the gastrointestinal (GI) tract has been widely recognized, most workers have assumed that enough of the active m aterial w ill always be either soluble or solubilized in the body and subsequently eliminated in the urine to permit urinalysis to be used reliably for the routine detection of internal contamination. Even in the soluble form,, many elements are still eliminated predominantly by way of the GI tract. The increased use of in vivo whole- body counters in recent years for the detection of gamma-emitting nuclides in human subjects has demonstrated conclusively that analysis of urine is not adequate as a general technique for the detection of internal contaminants. The point of emphasis is not only that the radionuclides will not be present in the urine, but also that such a high'percentage of the radioactive contami­ nants to be encountered are likely to be in the form of insoluble particulates, especially around operating reactors. This paper presents the experience gained in over three years of whole-body counting at the National Reactor Testing Station (NRTS) in the United States which encompasses perhaps the largest concentration of operating reactors of different types to be found anywhere in the western world. Thirty-one different radionuclides have been identified in human subjects from approximately 4000 whole-body examinations of nearly 2000 individuals. Where possible, the effective half-lives and modes of elimination of the various nuclides have been determined. Finally, a portable and relatively inexpensive counter has been developed so that most laboratories can take advantage of the sensitivity and freedom from many of the problems associated with excretion analysis that can be obtained by direct in vivo counting of human subjects.

2. INSTRUMENTATION

The whole-body counter employs a shield of pre-atomic era, armour plate steel, 11 in thick. The shield is in the form of a cylinder, 54 indiam., 94 in in length, and 5.5 in thick, with an additional flat piece 5.5 in thick added to the top, bottom and both sides. Both ends of the cylinder are closed with a double layer of the 5.5-in plate with the plates on one end being hinged to form an 11-t door. The inside of the cylinder is lined with 1/4 in of lead. The detector consists of a 4-in thick by 8-in diam. thallium-activated sodium iodide crystal optically coupled to three matched multiplier photo­ tubes. To minimize backscatter, the assembly is mounted near the centre of the counting shield but can be programmed to scan the subject in two directions. The output of the detector is fed into a 400-channel transistor­ ized pulse-height analyser, and the resulting spectra can be either printed out digitally by an IBM typew riter, plotted graphically on an X -Y recorder, or punched out on paper tape. Actually, all three methods are used: the graph for visual examination, the digital printout (which is printed on the same piece of graph paper containing the plot) for easy filing for a permanent record, and the paper tape for analysis. The punch tape is read directly 220 С. W. SILL et a l. into a high-speed electronic computer for data reduction. To make the counting rate as independent as possible of the distribution of the radionuclide within the body, the subject reclines in a chair that positions the body on a 50-cm arc around the detector with the legs below the knee remaining horizontal for comfort. The head is also tilted backward slightly to decrease the shielding of the thyroid gland by the chin. Two new counters with cubical shields approximately 9 ft on each inside dimension are presently nearing completion to replace the one described a b ove.

3. COUNTING PROCEDURE .

Experience has shown that it is generally not necessary for the subjects to take a shower before being counted for routine surveillance. This mini­ mizes the time lost by each individual from his normal work. The subject is required to change from street clothes to a disposable paper examination gown to prevent possible contamination of the counter. The disposable paper gowns are m ore convenient to keep on hand in quantity and less expensive to use than cotton coveralls. The analysis usually shows nothing abnormal in the spectrum so that the data are recorded and the subject released. In the small percentage of cases in which foreign activity is detected, the data is discarded, and the subject is required to take a shower with particular attention being given to cleaning the hair, hands and fingernails. The analysis is then repeated and the data recorded. The time required to repeat an oc­ casional 10-min count after showers is completely negligible in comparison to the time savedby eliminating showering as a routine part of the examination. Obviously, a shower is always required prior to research work, or other low-level non-routine counting. The spectra are analysed by either manual or instrumental spectrum stripping or by computer techniques.

4. RESULTS

During the last three years a relatively large number of individuals with small but easily detected internal burdens of various radionuclides has been encountered at the NRTS. In some cases the exposures resulted from known incidents; in most, they resulted from inadvertent contamination during normal operations and were detected during routine whole-body counting. However, except for some of the exposures sustained during recovery oper­ ations after the ill-fated accident at the SL-1 reactor, all other cases of internal contamination have been so low as to have little or no physiological significance to the individuals involved. About 95% of all the nuclides en­ countered were below 0.1 /ас and only one was above the permissible level for the respective nuclide. However, even in the range from 0.1 /j c down to the detection limit of a few thousandths microcurie, valuable data could be obtained with respect to effective half-lives and modes of excretion. Ob­ viously, actual data on human subjects even at such low levels are more valuable than sim ilar conclusions derived by extrapolation from animal data. Because of the circumstances related to the incidents, the principal mode of contamination in most cases was undoubtedly by inhalation, but very little is known about particle sizes or chemical or physical form. Each individual COMPARISON EXCRETION ANALYSIS WITH WHOLE-BODY COUNTING 221 containing a radionuclide of sufficient activity to be of interest was analysed repeatedly by whole-body gamma spectrometry until either the activity was completely gone or until the desired information had been obtained. Im­ mediately after the initial detection of the internal burden appropriate samples of urine and faeces were collected for ànalysis. Twenty-four-hour urine samples were evaporated to about 75 ml and counted for 5 min in a 3-in X 3-in thallium-activated well counter for routine screening. When in detectable quantity, the activity. was identified by gamma spectrometry, using a 400-channel pulse-height analyser in conjunction with chemical separations when necessary. Faecal samples were treated similarly.

4.1. The SL-1 incident

On 3 January, 1961, a nuclear excursion occurred at the Stationary Low Power Reactor No. 1, known as the SL-1, resulting in death to the three-man crew and extensive damage to the reactor core. Fission-product contami­ nation was extensive only inside the reactor building. Even though respira­ tors and other protective equipment were used, small internal exposures were received by the recovery teams and during the subsequent dismantling and clean-up of the reactor. All personnel entering the SL-1 area were monitored for internal contamination by gross gamma counting of urine. The major isotopes found both by whole-body counting and in urine samples of personnel making the first entries into the reactor building were primarily I131, with small quantities of Cs137 and Ba-La140. Caesium-134 was also ob­ served in some of the personnel involved in the subsequent dismantling and clean-up operations. The effective half-life obtained from urinalysis data for I131 was approximately 7.4 d. Sixteen individuals received an infinity dose to the thyroid in excess of 60 mrad. Only six exceeded 200 mrad and the maximum value was 1.1 rad. The Sr90 dose to the bone was also calculated from urinalysis data, using a power function model. The three highest individuals received doses of 10, 2 and 1 mrad for the first year. Unfortu­ nately, due to the nature of the emergency, no faecal samples were analysed. Retention curves for Cs137 obtained by whole-body counting of eight dif­ ferent subjects gave effective halfLlives varying from 90 to, 281 d. Six of the values were in the range of 90 to 128 d in comparison to the values re­ commended by the International Commission on Radiological Protection (ICRP) of 70 d for the total body and 138 d for muscle. Assuming muscle to be the critical organ, the infinity dose for the individual of highest activity was 32 mrad. Although later experience has shown that caesium is elim i­ nated partially by way of the gastrointestinal tract, the primary mode appears to be by way of the urine. Caesium-134 has also been observed 361 times in 168 individuals with a maximum body burden of 0.14 ц с.

4.2. Silver-110m

An experimental loop containing silver-soldered thermocouples in the Engineering Test Reactor (ETR) contained a significant quantity of the acti­ vation product Ag110m which was released into the reactor building when the loop was opened. Approximately 50 people received a small internal contamination initially which undoubtedly occurred through inhalation. Some 222 C .W . SILL et al. re-exposure occurred during subsequent clean-up operations which com­ plicated interpretation of the data obtained. Maximum activity was observed in most cases in the area of the nose and mouth, the chest, and the lower edge of the rib cage, and was found in nose wipes and sputum samples. In one individual a marked decrease in body burden was noted from whole-body counting after extensive coughing due to a cold. The primary mode of elim i­ nation was by way of the GI tract with so little voided in the urine that in almost all cases it could not be detected even in a 1500-ml sample. Data on the subject counted for the longest period of time before recontamination occurred gave an effective half-life of about 8 d. However, the individual showing the highest activity who was involved in the incident had an initial body burden of approximately 0.93 дс and an elimination rate corresponding to a half-life of approximately 13 d. Two other individuals exhibited effective h alf-lives of 17 and 69 d. None of the observed values agrees w ell with the value of 4.9 d given by the ICRP for the total body. However, the source of the radionuclide and the time of exposure were not the same for all the subjects. The initial whole-body counts were taken early enough after ex­ posure in most cases to detect the rapid clearance that occurs in the early part of the curve. The results indicate the variability in the elimination rate that occurs with insoluble m aterial in the lungs. Including the above incident, Ag-110m has been identified 583 times on 186 different individuals.

4. 3. Cobalt-58 and 60

Except for I131 and Cs137, Cc£° has been encountered more frequently during the past three years than any other nuclide, being identified by whole- body counting in 791 analyses on approximately 333 individuals. The highest burden was 1.5 дс and was received by an individual who removed his respir­ ator momentarily while working inside one of the large casks used for the shipment of radioactive materials. The activity, was detected on nose wipes ^ imm ediately after the incident which supports the conclusion that the exposure was due to inhalation. The activity in the body dropped abruptly to 0.21 дс in four days after which the elimination slowed up. Over a period of 11 months, the level dropped from 0.21 дс to 0.13 дс for an effective half-life of 2.0 yr. Essentially all of the activity was voided in the faeces which is very signi­ ficant since soluble cobalt is known to be excreted almost completely by way of the urine. The first two urine samples collected during the initial abrupt drop in body content contained some Co60 but it was less than 1% of that voided in the faeces for the same period of time. These were the only times that Co60 has been found in urine, even though one other individual also had a body burden as high as 1.5 дс. However, in this case his body burden dropped to less than 0.1 дс in six days after which he became recontaminated. Ex­ cretion curves on four other individuals followed fo r about two months showed half-lives ranging from 70 to 177 d compared to the ICRP value of 9.5 d for the total body. Cobalt-58 has also been observed 62 times on 50 different individuals with a maximum value of 0.03 дс. COMPARISON EXCRETION ANALYSIS WITH WHOLE-BODY COUNTING 223

4.4. Zinc-65

During the removal of a metal liner from an experimental loop in the Materials Testing Reactor (MTR), Zn65 was released into the work area. The activity was believed to have resulted from the activation of an insulating material containing stable zinc which was packed around the liner. Although 48 people were exposed, the highest value was only 1.2 цс, or about 2% of the maximum permissible body burden of 60 цс recommended by the ICRP. Eight of the individuals contained sufficient activity to follow for at least 100 d post exposure. The effective half-lives calculated from the straight-line portion of the retention curves ranged from 85 to 134 d. These values are all much shorter than the .194-d value given by the ICRP. The activity could be de­ tected in faeces as long as any significant body burden remained but could not be detected evèn in 24-h urine samples collected immediately after the incident. Although soluble zinc is normally eliminated predominantly in the faeces, even 10% eliminated in the urine, if present in soluble form, would be easily detectable. It was detected in urine only once in this entire investigation on a 24-h sample collected 60 d after exposure from an individual having a body burden of 0.39 juc at the time of sampling. Over­ all, Zn65 has been identified 505 times in 171 different people.

4. 5. A R E A hot c e ll incident

Six people were exposed accidentally in the hot cell in the Arm y Reactors Experimental Area (AREA) to a mixture containing primarily Ce141' 144, Ru103"106, Zr-Nb95 and Ba-La140. The highest body burden was approximately 1.7 дс of Zr-Nb95. Beginning immediately after the initial whole-body ana­ lysis, all urine and faeces were collected for several days from the two individuals having the highest burdens. Elimination of the activity from the body was very rapid and, with one exception, occurred exclusively by way of the GI tract. In fact, one of the individuals with the highest body burden eliminated essentially all of his internal burden in a single faecal excretion, as shown by whole-body counts taken immediately before and after the sample was given, all in the space of about 2 h. The other person with a high body burden eliminated the activity more slowly and could be followed easily for about three'days. Gamma spectra of faecal samples from both individuals were identical qualitatively to those obtained for the whole body and a smear taken from the work area in which the contamination occurred showed that no detectable fractionation of the different components had taken place during passage through the body. It is particularly informative to note that a faecal sample collected 45 h after the exposure contained several times as much activity as one taken only 22 h afterwards. Obviously, the highest activity will not necessarily be found in the samples taken earliest even though one day may have elapsed since exposure.. None of the nuclides could be de­ tected even in 24-h urine samples, except in one sample taken on the third day after the incident. The activity was shown by gamma spectrometry to be pure Ru103"106 with no trace of the other components of the mixture. Even the rutheniums were not detectable again in urine either before or after the third day. 224 С . W. SILL et al.

4. 6. Manganese-54 ■

Although Mn54 has been detected 98 times in 51 different individuals, the maximum value observed was 0.16 дс. Even so, the elimination was followed easily by whole-body counting for 84 d, giving a fairly smooth curve with an effective half-life of 64 d in comparison with the ICRP value of 5.6 d for the total body. The activity was readily detected in faecal excretion but could not be detected in 24-h urine samples, even though special techniques were used to increase the sensitivity.

4. 7. M ercu ry -1 9 7 and 203

Mercury-203 has been encountered 28 times in six different individuals with a maximum burden of 0.16 ц с. The individual with the highest burden was measured by body counting over a period of 92 d. Another person having an initial burden of 0.054 цс was measured five times over a period of 34 d. The data from both individuals gave an excellent fit to a straight line on a semi-log plot with a half-life of 25 d in comparison with the ICRP value of 8.2 d for the total body. The nuclide was eliminated in both urine and faeces, with the latter predominating. This nuclide is one of the very few that was detected initially through a routine urinalysis. Significantly, the ratio of faecal to urinary excretion was about 1.5 and was almost identical 28 d after the exposure, as it was during the first few days. M ercury-197 has been seen seven times in three different individuals with maximum values of 0.7 цс, 0.45 цс, and 0.29 цс. Only the individual with the lowest body burden was available subsequently to permit deter­ mination of the characteristics of this nuclide. However, five whole-body analyses in a 6-d period gave an effective half-life of two days in com­ parison to the ICRP value of 2.1 d. The ratio of faecal to urinary excretion was identical to that found fo r Hg2°3. 0 4.8. Plutonium incident

During modification of a dry box that had previously been contaminated while welding stainless-steel plutonium capsules, a number of individuals were exposed to oxides of Pu239 and/or Pu240 during and after opening the box. A total of 99 urine samples from 78 individuals were analysed with only one statistically positive result from each of six people. The results ranged from a high of 0.18 disintegrations/min ml to just barely above the detection limit of 0.0004 disintegrations/min ml for 1000-ml samples. On the other hand, 68 faecal samples from 22 individuals gave positive results on ten of the people. Due to the preoccupation with urine samples to deter­ mine the potential dose to the bone, faecal samples were not taken until five to ten days after exposure. The first samples taken from the four indivi­ duals with the highest burdens contained 26, 10, 3.8, a n d 1.6 disintegrations/ min g of faeces. The faecal activities decreased by factors of 15 to 100 during the next ten days, after which the excretion curves showed effective half-lives of 15 to 30 d. The data were very erratic and cannot be inter­ preted too precisely. However, the conclusion is clear that there was very little elimination of plutonium in the urine from this incident, and the dose COMPARISON EXCRETION ANALYSIS WITH WHOLE-BODY COUNTING 225 to the bone might be considerably less important than that to the lung, or possibly to the lymph nodes.

4. 9. O th er n uclides

Other nuclides that have been identified in human subjects at the NRTS are listed below, with the number of times seen and the number of individuals involved, in that order: Cr51, (15) 10; Zr-Nb95, (4 2 7 ) 2 3 2; Ru103~i°6, (93) 75; Ba-La140 (90) 51; Ce141' 144, (59) 49; Та^г, (50) 36; P a 233 ц з ) ю ; Np239, (1)1; Mo-Тс", (8)5; Sb122'125, (5)5; I132, (8)7; P 33, (3)3; and T e ^ 2, (6) 6. An effective half-life of slightly over one day was obtained from one of the Mo-Т с" subjects in comparison with the ICRP value of 1.8 d.

5. DISCUSSION AND CONCLUSIONS

In three years of whole-body counting of human subjects at the NRTS, over 95% of all the nuclides encountered could not be detected in 1500-ml samples of urine, the mode of elimination being almost exclusively by way of the GI tract. In many cases sufficient activity was eliminated from the body to have been easily detected in the urine even with those elements such as zinc and manganese that are normally eliminated predominantly in the faeces. Consequently most of the exposures encountered must have been sustained through the inhalation of insoluble particulates. With the exception of iodine and a few other special cases, insoluble particles may also be expected to be the most likely source of internal contaminants to be en­ countered at other installations, particularly those involving nuclear reactors. Particles already only sparingly soluble can be activated by the high neutron fluxes available, or radioactive materials can be converted into refractory particles by the high temperatures and oxidizing conditions in the presence of moisture that are so frequently present. Except for iodine, by far the largest number of exposures occurring at this site has been to activation products rather than to fission products, and maintenance men have been contaminated as frequently as reactor workers. When a relatively soluble element such as cobalt is not detectable in a 24-h urine sample from a body containing 1.5 цc, it is evident that the metabolic fate must be more dependent on the physical characteristics of the particle with which the nuclide is as­ sociated than with the ionic chemistry of the element itself. . Urinalysis cannot be recommended as a general technique for the de­ tection of internal contaminants, but is of value in certain cases. First, in every case in which I131, Hg197and Hg203, Cs134andCs137, and M o-Tc99 have been detected by whole-body counting, the same isotopes were present in easily detectable, although not predominant, quantities in the urine. For these elements urinalysis is not only adequate but is the preferred method because of the greater cost of whole-body counting. This is particularly fortunate in the case of iodine since the hazard from I131 is frequently the limiting factor during tests involving experimental reactors, and numerous analyses are required. Secondly, any elements that are soluble - and re­ main so under the prevailing conditions - might be expected to be absorbed into the blood-stream and eliminated at least in part by way of the urine. •However, insufficient attention has been given in the past to the chemical 226 С . W. SILL et al. characteristics ofthe various elements. Even in carrier-free tracer quantities, ter- and quadrivalent ions such as zirconium, thorium, plutonium, etc., hydrolyse rapidly under the mild conditions found in the lung. The insoluble compounds produced would then be eliminated primarily through the GI tract like other insoluble materials of comparable particle size and characteristics. On the other hand, soluble salts of uranium, cobalt, e tc ., do not hydrolyse readily and would be expected to be excreted predominantly in the urine. Third, urinalysis is undoubtedly useful for the determination of internal dose from nuclides systemically deposited in organs other than the lungs and GI tract. ' If the element is already fixed in the organ, it is difficult to see how significant elimination could be accomplished other than by trans­ location through the blood stream and elimination at least partially in the urine. ’ Two very important points should be emphasized concerning the chemical analysis of faeces. First, because of the high calcium content the addition of sulphuric acid or sulphates should be avoided; otherwise, the resulting calcium sulphate will occlude significant quantities of most ter- and quadri­ valent elements with ionic radii larger than about 1Â unit, such as thorium, rare earths and the transuranium elements. Second, samples should not be dry-ashed, or even allowed to dry and bake during wet ashing, unless the residue is fused with pyrosulphate. Otherwise the phosphates present are converted to a white solid that is extremely insoluble even in concentrated acids and will contain most of the components present in the sample.

6. PORTABLE WHOLE-BODY COUNTER

Although less convenient and more difficult to interpret, the analysis of faeces is much more sensitive and inclusive as a general technique for the detection of internal contaminants than the analysis of urine. For gamma - emitters, whole-body counting is even more sensitive and precise than faecal analysis without the disadvantages of the latter. The entire sample is taken for analysis and the activity present in the body is determined directly rather than having to be inferred from the activity that has been eliminated. How­ ever, the high cost and immobility of the heavy shields generally used to obtain very low backgrounds are distinct disadvantages that have prevented the more widespread use of this very sophisticated technique. Although very low backgrounds are necessary for the most refined work, they are not necessary to detect most gamma-emitting nuclides at concentrations well below their maximum permissible levels. In the opinion of the authors it is preferable to give up a little sensitivity in order to gain the reliability af­ forded by whole-body counting than to continue using urinalysis for per­ sonnel monitoring when it is not giving the general protection usually supposed. To prove that heavy shielding is not necessary for practical personnel protection, a portable whole-body counter weighing only 650 lb has been built and tested in this laboratory. The prototype is shown in Fig. 1. The frame is made of aluminium tubing and the only shielding employed for the subject is a 1/2-in sheet of lead under the seat cushion and in the upholstered back and sides of the chair. A 3-in X 3-in Nal(Tl) crystal with an integrally- coupled multiplier phototube is mounted in a 3/4-in lead shield which looks into the lead pocket created by the shielding in the chair. The end of the COMPARISON EXCRETION ANALYSIS WITH WHOLE-BODY COUNTING 227

■ . Fig.l . '

A portable whole-body counter

shield is flared out from the crystal to permit an angle of view of about 80° to include most of the trunk when the end of the detector shield is in contact with the subject. The complete detector assembly weighs about 50 lb but is easily lifted by a hydraulic jack and adjusted by manual rotation in any direction. Under normal background conditions approximately 0.01 цс of Cs131 can be detected in a human subject in a 10-min count. With the instrument located within 100 yd of the MTR and the reactor operating at 40 MW, the sensitivity decreased less than a factor of ten due to the increased back­ ground. When compared to a maximum permissible body burden of 30 цс, the detection of even 0.1 ^c is still adequate sensitivity for practical personnel protection. The sensitivity will be similar for most gamma-emitting nu­ clides. However, a counter is being mounted in a small yan for routine use at the NRTS that will employ relatively more shielding for use in reactor areas but will still be a relatively "unshielded" counter by current standards. The cost of the counter is highly dependent on the auxiliary instrumentation desired. The-portable unit shown in the photograph cost about $21 500 with only about $1500 being required for the detector, chassis and "shield". The remaining cost is for the instrumentation employed which includes a 400-channel pulse-height analyser, X -Y recorder, paper tape punch and reader, and automatic typewriter (not shown). However, a complete instrument with a 100-channel analyser and printer can be obtained for about $7500. 228 C.W . SILL et a l.

ACKNOWLEDGEMENT

The authors acknowledge the contributions of their associates in the ID Health and Safety Laboratory. Special credit is given to DeRay Parker and P. R. Boren for the design and construction of the portable whole-body counter, to A. R. Harbertson and L . L . Skolil fo r some of the ea rly work on the instruments and in setting up the counting programme, andtoL.E. Howard for obtaining much of the data presented.

DISCUSSION

. F. J. B R AD LE Y: Have you any information on thé chem ical form of the radionuclides in your inhalation/ingestion exposures? C.W. SILL: No. Nearly all our experience derives from small acci­ dental incidents, resulting for example from maintenance work on reactor components. Consequently information on either chemical or physical characteristics is generally not available. ■ J.F. ROELS: Do you think that one can replace the whole-body counter by urinalysis and faecal analysis for emergency and routine analyses? C.W. SILL: In general, I do not think so. Urinalysis is completely inadequate for both routine work and emergency use, although it does have specialized uses. Faecal analyses are much more sensitive and inclusive for the general detection of internal contamination but they present serious and undesirable problems of sampling, analysis and interpretation. It is no longer necessary to lim it oneself to bioassay methods in view of the relatively inexpensive whole-body counters now available. The instrument described in my paper can be obtained for about $7500. A. KAUL: Have you any explanation for the unusual case of virtually total excretion of the whole-body burden with the faeces? And have you been able to compare quantitatively by whole¡-body counting the results before and after excretion with those obtained by direct analysis of the stool sample? C.W. S ILL: I do not have a good explanation for this total elimination in a single sample. I was quite surprised myself. The activity might have been present as a single particle or at any rate very few particles. Exposure was fairly recent and probably in rather large aggregates which were not taken very far into the lungs and, consequently, eliminated easily. We have frequently been able to obtain material balances within 5 to 10% between the. amount of activity found in the total excretion and that known to have left the body between two consecutive whole-body counts. . A. KAUL: But in this case it'might be difficult, since you have practi­ cally a point source within the large intestine. Did you calibrate your body counter for this special geometry? C.W. SILL: Yes, as well as we could. However, the distribution in the body - particularly in the trunk - does not affect the counting rate too significantly when the 50-cm arc is used. N. I. SAX: What rapid method do you employ for the analytical compu­ tation of your whole-body counts? . C.W. SILL: The data from the 400-channel spectrometer is put on punched paper tape and reduced by sequential subtraction of standard spectra COMPARISON EXCRETION ANALYSIS WITH WHOLE-BODY COUNTING 229

in a computer'programme, giving both qualitative identification and quanti­ tative determination. E. E. POCHIN: In the instance where the entire measured body content of various nuclides was lost in a single faecal excretion, was evidence ob­ tained that these nuclides were initially present in body tissues, or may they already have been in the gut when first measured? C.W. SILL: No specific proof is available either way, but the circum­ stances surrounding the incident indicate that it is most likely that the con­ tamination occurred by inhalation, followedby rapid clearance into the gastro­ intestinal tract and subsequent elimination in the faeces. It is highly un­ likely that there was any systemic deposition of significance in the short time that elapsed between exposure and elimination.

THE ROLE OF FAECAL ANALYSIS IN A BIOASSAY PROGRAMME

J. D. EAKINS AND A. MORGAN HEALTH PHYSICS AND MEDICAL DIVISION, ATOMIC ENERGY RESEARCH ESTABLISHMENT, HARWELL, ENGLAND

Abstract — Résumé — Аннотация — Resumen

THE ROLE OF FAECAL ANALYSIS IN A BIOASSAY PROGRAMME. The role of faecal sampling and analysis in a bioassay programme is discussed, with particular reference to the estimation of retained lung burdens of insoluble radioactive materials. Experience has shown that urine analysis alone cannot always be relied upon to give an adequate indication of exposure by inhalation of insoluble radioactive materials. The results obtained are affected by so many variables that they defy interpretation. The analysis of faecal samples, collected after an incident involving airborne contamination, can confirm whether or not a significant intake has occurred.and will enable an initial estimate of the retained lung burden to be made by reference to one of the models describing the retention and elimination of inhaled particles. The subsequent faecal excretion pattern and particle size measurements on air filter samples representing the inhaled aerosol (if available) can be used to modify the initial estimate. : At the Atomic Energy Research Establishment (AERE), Harwell, the sampling and analysis of faecal samples is used as a complement to urine analysis, following cases of known or suspected exposure by inhalation. This method is considered to be the only satisfactory way of detecting and assessing lung burdens of insoluble compounds of Pu233, which cannot be detected with adequate sensitivity by in vivo counting. Some examples of excretion patterns obtained from cases of accidental inhalation of insoluble compounds of plutonium and thulium are described. •

IMPORTANCE DE L'ANALYSE DES MATIERES FÉCALES DANS UN PROGRAMME D’ANALYSES BIOLOGI­ QUES. Les auteurs étudient l'importance du prélèvement et de l'analyse d'échantillons de matières fécales dans un programme d'analyses biologiques, plus particulièrement du point de vue de l'évaluation des charges de matières radioactives insolubles qui auraient été retenues dans les poumons. L'expérience montre que l'on ne peut pas toujours s'en remettre à la seule analyse d'urine pour dépister de façon assez sûre les radio­ expositions par inhalation de matières radioactives insolubles. Les résultats obtenus sont influencés par un si grand nombre de variables qu'ils défient toute interprétation. L'analyse d'échantillons de matières fécales prélevés après un incident entraînant contamination de l'air ambiant, peut corroborer qu'il y a eu ou non absorption de quantités significatives et permettra de faire une première évaluation de la charge retenue par les poumons, grâce à l'un des modèles qui représentent la rétention et l'élimination des particules inhalées. Pour modifier l'estimation initiale, on peut se fonder sur le régime des excrétions ultérieures en fonction du temps et sur les mesures de la dimension des particules dans des échantillons représentatifs de l'aérosol inhalé (s'il en existe), recueillis sur un filtre à air. Au Centre de Harwell, le prélèvement et l'analyse d'échantillons de fèces servent à compléter les analyses d'urine, depuis qu’on a constaté ou soupçonné chez certains sujets une radioexposition par inhalation. On estime que cette méthode constitue le seul moyen satisfaisant de déceler et de mesurer les charges de

composés insolubles de 239 Pu dans les poumons, qui ne peuvent être décelées par la méthode du comptage in vivo dont la sensibilité est insuffisante. Le mémoire donne des exemples de régimes d'excrétion pour des sujets qui avaient inhalé accidentellement des composés insolubles de plutonium et de thulium.

ЗНАЧЕНИЕ АНАЛИЗА ЭКСКРЕМЕНТОВ В ПРОГРАММЕ БИОЛОГИЧЕСКИХ ИССЛЕДО­ ВАНИИ. Обсуждается роль взятия проб и анализа экскрементов в программе биологических исследований, особенно в отношении определения содержания осевших в легких нерастворимых радиоактивных материалов. Опыт показал, что не всегда можно полагаться на результаты исследования мочи,.как надежного показателя облучения после вдыхания нерастворимых радиоактивных материалов. На полученные данные влияют так много переменных величин, что они не поддаются интер­ претации. Однако анализ проб экскрементов, собранных после облучения, протекающего с

231 232 J. D. EAKINS and A. MORGAN

загрязнением воздуха, может показать, имело ли место значительное поглощение, и дать возможность предварительно определить осевшие в легких изотопы путем сравнения с одной из моделей, описывающих задержку и выделение вдыхаемых частиц. Последующее выделение с экскрементами и измерение размера частии в пробах с воздушного фильтра, представляющих вдыхаемую аэрозоль (если это возможно), могут быть использованы для модификации пред­ варительных результатов. В научно-исследовательском центре по атомной энергии в Харуэлле взятие образцов и анализ экскрементов используются в качестве дополнения к анализу мочи в случаях известного или предполагаемого облучения в результате вдыхания радиоактивных материалов. Этот метод считается единственным удовлетворительным способом обнаружения и оценки содержания в организме нерастворимых соединений Pu239 f который не может быть обнаружен с достаточной точностью путем подсчета in vivo. Дается описание некоторых примеров выделения, полученных после случайного вдыхания нерастворимых соединений тулия и п луто н и я. Установлено, что радиоизотопы некоторых элементов преимущественно удаляются с экскрементами независимо от пути поступления. При случайном облучении такими радио­ изотопами результаты только анализов мочи невозможно использовать в качестве показателя облучения, и при отсутствии других методов обнаружения следует произвести взятие проб экскрементов и их анализ. Описан случай, когда рана была загрязнена частицами облученного урана и тория. В последующем ассоциированные продукты деления и протактиний-233 вы­ делялись почти целиком с экскрементами.

EL PAPEL DEL ANALISIS DE HECES EN UN PROGRAMA DE BIOANÁLISIS. La memoria examina el papel que corresponde al muestreo y al análisis de heces en un programa de bioanáliáis, principalmente para evaluar las cargas de sustancias radiactivas insolubles retenidas en el pulmón. La experiencia demuestra que no siempre se pueden considerar los resultados del análisis de orina como indicio seguro de que ha habido exposición por inhalación de sustancias radiactivas insolubles. En los resultados obtenidos intervienen tantas variables que no es posible interpretarlos. El análisis de muestras de heces recogidas después de un accidente en que intervenga la contaminación arrastrada por el aire permite confirmar si se produjo o no una absorción significativa, asf como evaluar en primera aproximación la carga retenida en el pulmón basándose en uno de los modelos que describen la retención y eliminación de partículas inhaladas. Esta evaluación inicial se puede corregir ulteriormente al establecer el esquema de excreción por las heces y, si se dispone de ellos, aplicando los resultados de las mediciones de los tamaños de partícula de muestras representativas del aerosol inhalado, tomadas con un filtro de aire. En los Laboratorios de Harwell, se practica el muestreo y análisis de muestras de heces para complementar el.análisis de orina en aquellos casos en que se sabe o se sospecha que han sido inhaladas sustancias radiactivas insolubles. Se considera que este método es e l ún ico satisfactorio para detectar y evaluar la carga pulmonar de

compuestos insolubles de 2 3 9 Pu, que no pueden detectarse con la sensibilidad apropiada por recuento in vivo. Se exponen algunos ejemplos de esquemas de excreción obtenidos en casos de inhalación accidental de com­ puestos insolubles de plutonio y de tulio.

1. INTRODUCTION

In the Atomic Energy industry the analysis of urine samples has been used for many years as a means of detecting internal contamination with radioactive material. The detection of a radionuclide in urine is evidence of _a systemic burden and in circumstances where people are exposed to soluble compounds of radionuclides with well-established renal excretion patterns, the analysis of urine samples can provide a sensitive indication of internal contamination. In addition, if the relationship between retention and ex­ cretion has been established by human experimental studies, then the body burden may be assessed from the results of urine analysis, with an accuracy limited only by biological variations from one individual to another. FAECAL ANALYSIS IN BIOASSAY 233

Failure to detect a radionuclide in urine does not,however, denote the absence of radioactive material in the body. SILL [l],for example, has listed several radionuclides (including C r51, Co60, Zn65, Z r96, Ag110m and Sb125) which had been identified in the body by in vivo counting, but could not be detected in urine, even in 24-h samples. It is becoming increasingly evident that, although urine analysis will provide a valid index of exposure to certain radionuclides under well-defined conditions, there are circum­ stances in which it will give misleading information, or possibly no inform­ ation at all, regarding the severity of internal exposure.

The purpose of this paper is to indicate the type of situation where urine analysis cannot be relied upon, and to show under what circumstances faecal sampling and analysis can be of assistance and how the results of such m ea­ surements may be interpreted.

2. FAECAL EXCRETION

Material excreted from the body in faeces may be endogenous or exo­ genous in origin. Excretion of material which has been systemically bound is termed endogenous while the elimination of material which has never been absorbed is referred to as exogenous excretion. Langham's representation of the body as a doughnut is a useful concept. The hole in the middle of the doughnut represents the gastrointestinal tract and the lung is a recess in the hole. Material which has not passed the alveolar surface, or which has not been taken up from the gastrointestinal tract, has never been inside the doughnut and th erefore does not constitute a system ic burden.

2. 1. Endogenous faecal excretion

Some radionuclides (e.g., Zn65, Ce144, Po210, Ra226, Ac227) are known to be eliminated from the body mainly in faeces, even when present as a systemic burden. They enter the gastrointestinal tract either directly through the epithelium, or in biliary secretions from the liver. In such cases it is usual to find that, although it may be small, there is a significant urinary excretion and if the maximum permissible body burden (MPBB) is high enough and the analytical method sensitive enough, the analysis of urini. may still provide an adequate index of exposure.

2. 2. Exogenous faecal excretion

It is well known that while some elements (e. g. ,. rare earths and many heavy elements) pass through the gastrointestinal tract with scarcely any absorption, others exchange with the body pool. This exchange may be com­ plete, as in the case of sodium, 'and potassium, or it may only be partial, as with calcium-, strontium, magnesium and phosphate [2]. Experience has shown that exogenous faecal excretion frequently occurs after exposure to radioactive airborne contamination, particularly in cases where the radio­ nuclide is in the form of an 'insoluble1 compound or incorporated in an in­ 234 J. D. EAKINS and A. MORGAN soluble matrix. Under the physiological conditions which exist in the body, it is doubtful if any m aterial can be regarded as 'in solu b le1 in the chemical sense and animal experiments with silver iodide [3] and strontium sulphatè [4], which are normally regarded as insoluble, have shown that they are transferred quite rapidly from the lungs into the blood stream. . With some materials,however, the absorption may be so small that quite large lung burdens of radioactive material may give rise to insignificant concentrations in urine, which cannot therefore be used as an index of exposure.

3. ASSESSMENT OF LUNG BURDENS OF INSOLUBLE MATERIALS

3. 1. In vivo counting

In vivo counting may bé used to detect lung burdens of gamma-emitters, beta-emitters with E^^XD.5 MeV (by measuring the associated bremsstrah­ lung) and alpha-emitters which give rise to X- or gamma rays in sufficient abundance. If whole-body counting equipment is available, then in vivo counting provides the most satisfactory method for the determination of lung burdens of most radioactive materials. There remains a few alpha-emitters which cannot be detected with adequate sensitivity by this method and the most important of these is Pu239. Even with techniques specially developed for the measurement of this radionuclide [5], the limit of detection is un­ likely to be adequate to detect a maximum permissible lung burden of 16 nc.

3. 2. U rin e analysis

Urinary excretion data can only be related directly to systemicallybound material and to use them to estimate lung burdens, some model describing the transfer of material from the lung into the blood-stream must be proposed. LANGHAM [6] and HEALY [7 J have suggested that the results of urine analysis can be used to estimate lung burdens of plutonium. Langham used his own human experimental data to estimate the systemic burden from the results of urine analysis and multiplied this by ten to obtain the lung burden. This factor was based on inhalation experiments with rats carried out by ABRAMS et al. [8] who found that ten days after exposure to aerosols of Pu02, about 10% had been absorbed from the lungs. Healy assumed that plutonium is removed from the lungs in an expo­ nential manner and that the overall rate of elimination can be regarded as consisting of two components, one describing solubilization and transfer to blood (Xs) and the other describing removal by ciliary action (Xc). By using Langham's equation relating retention and urinary excretion, Healy produced an expression involving Xs, relating urinary excretion and lung burden. • Inhalation studies made by BAIR et al. [9] using beagle dogs, have shown that urinary excretion of plutonium following inhalation of Pu02 is inversely related to particle size, so that as the particle size diminishes, more plu­ tonium is excreted in urine. Dogs exposed to the smallest particle size FAECAL ANALYSIS IN BIOASSAY 235

(M M D 0.23 /лm)* excreted plutonium at much greater rates than dogs ex­ posed to aerosols with larger oxide particles and even to soluble plutonium nitrate aerosols. It is evident that the transfer of radioactive material from the lung to the blood stream depends very much upon its solubility and while some ma­ terials considered 'insoluble' in the chemical sense are transferred quite rapidly, others such as PUO2 appear to dissolve only very slowly. In view of the present lack of experimental data on this subject, any attempt to re­ late urinary excretion levels to lung burdens must be regarded as hypothetical1.

3. 3. Faecal analysis .

To interpret the results of faecal analysis, some model describing the fate of insoluble particles deposited in the lung must be adopted. It is known that the fate of inhaled aerosols depends upon a large number of factors, including particle size, shape, density and chemical form. While the de­ position of aerosols of known size, shape and density can be predicted to some extent, details of these parameters are not always available in prac­ tice. The ICRP has produced therefore, a generalized lung model for use in situations in which specific data are lacking. In this model it is assumed that 75% of the inhaled particles are initially deposited in the respiratory system and that 25% are exhaled. Of the particles retained, two thirds are deposited in the nasal passages and epithelium of the bronchi, and are re­ moved from these sites to the throat by ciliary action, from where they are swallowed. It has been shown that cilia can move particles up the bronchi and bronchioles at 0.15 - 1.0 cm/min and up the trachea at 3 - 4 cm/min, so these particles are removed quite rapidly from the .respiratory system. The remaining third of the retained particles (representing the smaller sizes) will be deposited in the lower parts of the lung. The ICRP postulate that half of these will be elevated and swallowed within 24 h, while the other half is taken up into body fluids and is removed from the lung fairly slowly, with a h alf-life of 120 d. ** A different interpretation of the fate of insoluble particles deposited in the lower parts of the lung has been put forward by LANGHAM [6]. He suggests that 60% are phagocytosed or otherwise removed up the bronchial tree and then eliminated via the gastrointestinal tract with a half-time of about 180 d, and the remaining 40% are taken up through the alveolar wall, into the systemic circulation, with a half-time of a few days at most. The differences between these two models are illustrated in Fig. 1. A comparison shows that in each case the material deposited in the upper respiratory tract will be transferred rapidly to the throat and swallowed. In the Langham model,however, material initially deposited in the alveoli will be removed by a similar mechanism over a long period of time, so that faecal excretion will continue, albeit at a much lower rate for a considerable period. This is in agreement with the results of animal experiments in­ volving Pu02 aerosols [9, 10] which show that active lung clearance, involving

* MMD is the mass m edian diam eter.^ . * * Exceptions are plutonium and thorium, for which half-lives of 1 yr and 4 yr respectively are recommended. 236 J. D. EAKINS and A. MORGAN

DEPOSITION

lO O X Ii EXHALED 25%

Upper Respiratory Tract 50X

Lower Respiratory Tract 25%

Fig. 1

Schematic representations of lung models the ciliated bronchial mucosa, continues over many months. Whichever model one adopts,however, it is evident that a large proportion of-the in­ soluble material deposited in the lung will be removed fairly rapidly from the respiratory passages to the throat and swallowed. If the subsequent FAECAL ANALYSIS IN BIOASSAY 237 uptake from the gastrointestinal tract is low (i. e., if f-, is small) then the bulk of this material will be eliminated in faeces over a period of a few days. There is little human experimental work which enables faecal excretion rates to be related to retained lung burden following the inhalation of in­ soluble compounds. The experiments of BAIR et al. [9] with beagle dogs probably provide the only data which can be extrapolated to man with any confidence. These workers showed that, except for very small aerosols of РиОг (MMD 0.23 |jm), this material was excreted predominantly in faeces. In almost all their experiments with PuC>2, the faecal excretion rate on the tenth day after exposure fell in the range 0.2 - 1% of the body burden. It is evident that the analysis of faecal samples, following the suspected inhalation of insoluble radioactive particles, will not only confirm whether or not inhalation has in fact occurred, but will also enable a prediction, of the retained lung burden to be made on the basis of one or other of the lung models. In the following section a number of examples are given, in which faecal analysis has been used to detect the inhalation of insoluble radioactive materials and to assess the retained lung burdens. The examples also il­ lustrate the unreliability of urine analysis as a valid index of this type of exposure.

4. CASE D A TA

4. 1. Case A

This investigation arose following an incident in which a man (subject A) was exposed to airborne contamination escaping from a glove box through a split glove. Contamination levels of about 20 cps were found on his per­ sonal gloves and air samples, taken at distances of 4 and 12 ft from the box, showed 3 nc and 0.4 nc of alpha activity respectively. After the incident A blew his nose on a paper tissue which was wet ashed and found to contain 5.3 nc of alpha activity. This sample and the air filters were examined by alpha spectrometry and the activity was found to have the same composition, nam ely Pu239 and Am 241 in a 7:1 ratio. Autoradiographic and m icroscopic measurements on the air filters showed that only 1.3% of the activity was in the size range 0.2 - 1 ;um, 4% was on particles less than 2.5 (im and 15% on particles less than 5 (im. Most of the activity was associated with particles in the 5-13 цт range which would probably be trapped in the nose. Because of the high density of plutonium, these particles are dynamically equivalent to much larger unit density particles. An in vivo measurement of A, made on the fifth day, with a 9-in diam. Nal crystal at the back of the chest, showed the possible presence of Am241 in the lungs (not m ore than 1.3 nc), but this was not confirmed by subsequent measurements. A total of 6 nc of Pu239 was excreted by A in faeces in the 48 h after intake and subsequently levels rapidly diminished to a few pc/d. Significant levels of Pu239 were detected in urine for a period of about two weeks and the presence of a systemic burden was also confirmed by mea­ surements in blood. Blood samples taken on day 7 and day 18 gave positive results of 0.014 and 0.005 pc/ml, but subsequent samples taken on days 25 2 38 J. D. EAKINS and A. MORGAN

Fig-2

Faecal and urinary excretion following inhalation of insoluble plutonium (Case A) and 31 were below the limit of detection (<0.001 pc/ml). The urine and faecal excretion patterns are shown in Fig. 2. The in vivo measurement of Am241 enables an absolute upper lim it to the lung burden to be set. This measurement would certainly have detected 2.5 nc which, in this case, would be associated with about 20 nc of Pu239 (about one maximum permissible lung burden). From the air sample mea­ surements and a knowledge of the particle size distribution, SHERWOOD and STEVENS [11] suggest that A could have inhaled 15 nc of Pu239, but that only 250 pc would have been deposited in the deep lung. This estimate of the total activity inhaled is in good agreement with the sum of that removed from A 's nose and excreted in faeces (11 nc). The particle size distribution indicates that little of the activity could have penetrated to the alveoli, so that the faecal excretion should fall off rapidly and, as shown in Fig. 2, this was the case. According to the ICRP model, 62^-% of the inhaled material is swallowed rapidly and only 12|-% is •retained in the lung with a long half-life. Applying this model to A 1 s data gives a retained lung burden of 1.2 nc, but in view of the particle size d istri­ FAECAL ANALYSIS IN BIOASSAY 239 bution, this is likely to be an over-estimate. Applying Bair's data for fae­ cal excretion of Pu02 from beagle dogs gives a lung content of 0.2 - 1.0 nc at 10 d after intake. Assessing the results of faecal analysis in this way enables the re­ tained lung burden to be placed at probably not more than 1 nc of Pu239. As shown in Fig. 2 the level of Pu239 in urine is only significant (>0.025 pc/d) during the two weeks following intake. The highest urinary excretion oc­ curred on day 3, and similar delays in peak urinary output have been ob­ served in other cases of inhalation. This suggests that most of the systemic burden may be acquired.by absorption during the passage of material through the gastrointestinal tract, rather than through the lung wall. If this is the case, then there is even less justification for relating urinary excretion' levels to the lung burden.

4. 2. Case В

In this incident, a man (subject B) was allowing a specimen of plutonium wire, sealed in a double brass container, to warm up to room temperature. This specimen had been cooled in liquid helium for about a year and as the temperature increased the container exploded. On discovering that the plu­ tonium wire was missing, В immediately left the laboratory. Fairly high contamination levels were found on his clothing, face and hair and about 1 nc of Pu239 was rem oved from his nose on nasal swabs. A sputum sample collected the following morning showed no detectable activity. When the plutonium wire was located and weighed it was found that about 100 mg was missing. In an attempt to obtain information on the particle size distribution of the inhaled material, some of the contamination onB's clothing was exa­ mined and it appeared that only about 1% of the activity was associated with particles in the respirable region. After the incident, the medical staff prescribed magnesium sulphate as a laxative and, as shown in Fig. 3, rapid faecal elimination of about 1 nc of Pu239 occurréd in the first 24 h. Signifi­ cant levels of Pu239 in urine were detected during the first week after intake, but the highest daily excretion level did not exceed 0.1 pc. ' The faecal excretion pattern in this case suggests that little of the ac­ tive material penetrated to the deeper parts of the lung and that almost all would have been trapped in the nose and upper respiratory passages. Ap­ plying the ICRP lung model indicates that the retained lung burden was only about 200 pc and this is likely to be an over-estim ate.

4. 3. Case С

In this incident, a man (subject C) was posting drums containing active waste into a sealed waste disposal area and the operation gave rise to sig­ nificant levels of airborne contamination. Although air samplers were in operation during the period of exposure, the filters were not counted until several days had elapsed and complete urine and faecal sampling were not started until some time after the incident. Autoradiographic examination of the air filters showed that the contamination consisted of a mixture of 240 J. D. EAKINS and A. MORGAN

DAYS AFTER INTAKE

. Fig. 3

Faecal and urinary excretion following inhalation of insoluble plutonium (Case B)

Fig. 4

Faecal and urinary excretion following inhalation of insoluble plutonium (Case C) FAECAL ANALYSIS IN BIOASSAY 241

Pu02 particles (0.5 - 5 /ига) and contaminated dust particles of larger size (5 - 20 pm). About half the activity was associated with the Pu02 particles. The excretion curves for this case are shown in Fig. 4 and it is evident that there is a longer component in the faecal excretion pattern than in either case A or B. This must be attributed to the slow removal of the smaller particles initially deposited in the deeper parts of the lungs. The urine ex­ cretion pattern is confused by the fact that most of the activity in the first sample was identified as Po210 by alpha spectrometry. It is unlikely that the excretion of Pu239 in any of the urine samples examined exceeded 0.5 pc/d. As the faecal samples in the critical period following inhalation were not provided, it is impossible to assess the lung content, but the evi­ dence suggests that the retained lung burden may have been greater than in either of the two previous cases. This illustrates the importance of iden­ tifying a hazardous situation as rapidly as possible and the necessity, once a decision has been made, of collecting all faeces voided during the first few days following exposure.

4.4. Case D

Sub j ec t, D, bee ame heavily contaminated while .cle aning out a high - activity handling cell in which'were aluminium cans containing thulium-170 as the oxide. Surface contamination levels of 3000 cps. (0):were detected on his hands and 100 cps on his head and face. The installed a ir sam pler in the area indicated an air concentration of only 0.5 mpc, but was situated some distance from, the source of contamination. Subsequent measurements indi­ cated that about 100 цс of Tm1™ were removed from the laboratory via the extract system. . An in vivo measurement on the day after the incident showed internal contamination with Tm1™ and also Co60. Estimation of the lung burden at this time was complicated by the fact that much of the active material was still in the gastrointestinal tract. A second measurement on the fourth day, when most of the activity had been voided, indicated that about 20 nc of Tm170 and about 3 nc of Co60 w ere retained in the lung. Faecal and urine samples were collected from D for several weeks and the excretion patterns are shown in Fig. 5. Unfortunately the first faecal sample was not collected, but during the firs t week, o ver 650 nc of TmiTO w ere excreted in faeces, m ainly on days 2 and 3. The presen ce of Tm-t™ in urine was not detected at any time during the investigation, although the limit of detection was 10 pc. If the ICRP model is applied to the faecal ex­ cretion data, the retained lung burden would be about 130 nc, although in vivo counting gave a result of only 20 nc, considered to be accurate to within a factor of two either way. On the other hand, the analysis of urine would have given no indication of the intake.

CONCLUSIONS ■

Faecal sampling and analysis can be used as an, alternative to in vivo counting, to detect exposure to insoluble airborne contamination. An initial estimate of the retained lung burden of such materials can be made from 242 J. D. EAKINS and A. MORGAN

Ю 20 30 40 50

DAYS AFTER INTAKE. '

Fig- 5 .

Faecal and urinary excretion following inhalation of thulium oxide (Case D)

the results of faecal analysis, by reference to a lung model, provided that all'samples voided in the first few days after intake are available. The accuracy of this assessment will depend upon the particle size and solubility of the material. ■ 1 • . At AERE, Harwell, faecal sampling is only employed after known or suspected exposure, particularly to insoluble plutonium compounds, which cannot be detected in the lung with satisfactory sensitivity by in vivo counting. The decision to request faecal samples is taken at a fairly high lève! and care is taken to explain the reason to the persons involved and to keep them informed of the results and their significance. The operational health physicists have been made aware of the limitations of urine analysis and co-operate in bringing inhalation cases to the notice of the bioassay section, so that' complete sampling can be initiated at an early stage.

ACKNOWLEDGEMENTS

■ The authors would like to thank Messrs. D. Brown, P. Gomm, K. Groves and E. Tattersdill who carried out the analyses in the various cases des- FAECAL ANALYSIS IN BIOASSAY 243 cribed. They also acknowledge the full co-operation and assistance given by members of the Medical Services at Harwell.

■ REFERENCES

[1] SILL, C. W. , DP-831 (1962) 126. . • .

[2 ] COMAR, C . L. and BRONNER, F. , in: M ineral M etabolism _1A (BERGER, E. Y . , E d .) A cad em ic Press Inc, , New York (1960) 263. . ' . (3J WILLARD, D. H. and BAIR, W. J. , Acta Radiol. 55 (1961) 486. [4 ] BAIR, W .J. , WILLARD, D. H. , TEMPLE, L. A: and SMITH, D. G. , HW -5 9 5 0 0 (1959). [5] TAYLOR, В. T. and RUNDO, J. , AERE - R 4155 (1962).

[ 6 ] LANGHAM, W. H. , Brit. J. Radiol. Suppl. No. 7 (1957) 95. ■ [7] HEALY, J. W, , Am. Indust. Hygiene Assoc. J. 18_(1957) 261. . . .

[ 8 ] ABRAMS, R. , SEIBERT, H. C. , POTTS, A .M ., FORKER, L. L. , GREENBERG, D. , POSTEL, S.- and LOHR, W. , C H -3 6 5 5 (1 9 4 7 ). [9] BAIR, W.J. , TOMBRÓPOULOS, E. G. , and PARK, J. F. "Distribution and removal of transuranic elements and cerium deposited by the inhalation route", in: Diagnosis and Treatment of Radioactive Poisoning,

IAEA, Vienna (1963) 319. " , 1 [ 1 0 ] MORROW, P. E. and CASARETT, L. J. , in: Inhaled Particles and Vapours (DAVIES, C. N. , Ed. ) Pergamon Press (1 9 6 1 ) 167. . - [11] SHERWOOD, R J. and STEVENS, D. C. , Private Communication.

. ■ ' / DISC USSIO.N ■

L. M. SCOTT: Do you collect all faeces after a known exposure? And how do you assess total faecal excretion when only two or three samples a week are collected? , . A. MORGAN: The essential thing is to initiate complete collection of faecal excreta immediately. We usually continue this for a week, since with delays both in the excretion its e lf and in analysis of samples we do now know the magnitude of the problem ea rlie r. Thereafter sampling may be continued at appropriate intervals, or discontinued, as the situation demands. We are aware that excretion in the faeces is irregular, and have observed no corre­ lation between the amount of plutonium excreted and dry or wet weight of faeces. If collection is incomplete we have no better alternative than to assume each sample represents average daily elimination. After explaining the, need for faecal collection we get quite good co-operation from the indi­ viduals under examination. B. RAJEWSKY: I have two questions. Firstly, the curve for Case С (Fig. 4) is markedly different from the curves for Cases A, В and D (Figs. 2, 3 and 5). Why is this? ■ Secondly, does'the retention of particles depend on their size and elec­ trical charge, and did you measure these in your lung model? A. MORGAN: With régard to the excretion curve for Case C, this does indeed represent a pattern different from the three other cases which I cited as examples. I think the difference is due perhaps to the difference in particle size of the inhaled material. In the other cases I think that it is almost certain that the m aterial inhaled was mainly outside the respirable range and therefore collected or trapped in other parts of the lung and cleared very rapidly by the mechanism adopted by the ICRP in their lung model. Very 244 J. D. EAKINS and A. MORGAN

little material would have penetrated into the alveolar region and therefore there would be very little chronic elimination from the lung. Although we have no evidence, since the initial faecal samples were not collected, I suspect that in Case С the amount of m aterial inhaled was probably very much larger, and possibly the particle size also smaller, so that more material would have penetrated into the alveolae. The chronic excretion pattern we see here is'the slower mechanism of material being removed from the lower parts of the lungs. With regard to the effect of particle size, one seldom has ádequate measurements of particle size in cases of accidental exposure. If samples are taken they are usually taken at some distance from where the man was working. At Harwell we have only recently discovered, with the use of .personal air samplers in which the air-sam pling head is'actually on the man's coat just below his chin, that we were seriously underestimating the inhala­ tion exposure by assessing the results of air samples taken at some distance from the place of work. In a few cases we have had information on particle size and it'does seem that, where the particle size is large, elimination of plutonium from the lungs is very abrupt. Work has been done recently at H arw ell on the form of plutonium oxide particles. It appears that these exist in two form s, either as an oxide or as oxide particles attached to very much larger particles, and that the form of plutonium can vary from area to area at the Harwell establishment. I think that the abrupt excretion pattern occurs in cases where plutonium oxide particles are attached to larger p a rtic le s . I have no information on the charge of these particles. . B. RAJEWSKY: Our experience has shown that the la rger particles are mainly exhaled and do not penetrate deep into the lung. P articles of medium size penetrate very deep into the lung, while small particles remain in the nose and mouth. W.N. SAXBY: I should like to mention a case which has a bearing on those described by Dr. Morgan. Following intense contamination, by in­ halation, with insoluble airborne plutonium, the worker involved showed an acute plutonium excretion pattern sim ilar to those described by Morgan and others both in urine and faeces. Chronic plutonium excretion has continued at measurable levels over a period of three years in both urine and faeces, with the daily faecal excretion consistently exceeding that in the urine. RADIOCHEMICAL DETERMINATION OF PLUTONIUM FOR RADIOLOGICAL PURPOSES 4

J.M. NIELSEN AND T. M. BEASLEY* . CHEMICAL LABORATORY, HANFORD LABORATORIES, GENERAL ELECTRIC COMPANY, RICHLAND, WASHINGTON, USA

Abstract — Résumé — Аннотация — Resumen

RADIOCHEMICAL DETERMINATION OF PLUTONIUM FOR RADIOLOGICAL PURPOSES. In this paper the procedures that have been and are currently being used for the determination of micro- microgram quantities of plutonium in biological and environmental samples are reviewed. Special emphasis is placed on excretion analysis. Expected urinary excretion rates have been calculated, using assumed levels of plutonium deposition, so that the analytical sensitivities of various procedures can be compared. Complete dissolution of excreta, soil, bone, tissue and vegetation are described with emphasis on avoiding the formation of refractory com­ pounds of plutonium which are soluble with difficulty. Analytical methods for plutonium analysis of these materials are reviewed and include со-precipitation, liquid-liquid extraction, ion exchange chromatography and the use of plutonium isotopes for yield determinations by means of alpha energy analysis. Using counting statistics, comparisons are made of the sensitivities available in low-level alpha counting, using ionization chambers,- proportional counters, diode counters, and nuclear track emulsions. Isotopic analysis of plutonium by alpha spectrometry, nuclear emulsion techniques, and liquid scintillation counting are included. The use of non-isotopic carriers as a source of extraneous activity and the environmental levels of plutonium recently encountered around the world are discussed in connection with "blank" samples. Two possibilities are considered for future methods of plutonium analysis where increased sensitivity is required. These are activation analysis and fission fragment counting.

DETERMINATION RADIOCHIMIQUE DU PLUTONIUM A DES FINS RADIOLOGIQUES. Les auteurs exa­ minent Г ensemble des méthodes que l’on a employées et que Ton emploie actuellement pour évaluer les quantités de plutonium de l'ordre du picogramme présentés dans des échantillons biologiques et mésologiques. Ils insistent particulièrement sur l’analyse des excreta. Ils ont calculé des taux d’élimination urinaire pro­ bables à partir de concentrations de plutonium théoriques, de manière à pouvoir comparer'la sensibilité de différentes méthodes d’analyse. Les auteurs décrivent la dissolution complète d'échantillons d’excreta, de sols, d’os, de tissus et de végétaux en insistant sur la nécessité d'éviter la formation de composés réfractaires du plutonium qui sont difficilement solubles. Ils examinent les méthodes analytiques employées pour doser le plutonium contenu dans ces substances, notamment la coprécipitation, l'extraction par partage) la chromato- graphie sur échangeurs d’ions et l'emploi d'isotopes du plutonium pour la détermination des quantités produites par analyse de l'énergie des particules alpha. A l'aide de statistiques du comptage, les auteurs comparent les sensibilités obtenues dans le comptage de particules alpha de faible énergie au moyen de chambres d'ionisa­ tion, de compteurs proportionnels, de compteurs à diode et d’émulsions nucléaires. L’analyse isotopique du plutonium par spectrométrie alpha, les techniques .des émulsions nucléaires et le comptage par scintillateur liquide sont traités dans le mémoire. A propos des échantillons complètement inactifs, les auteurs étudient l'emploi d'entraîneurs non iso­ topiques en tant que source d'activité étrangère ainsi que les degrés de contamination du milieu ambiant par du plutonium récemment relevés dans le monde entier. Ils examinent deux possibilités comme futures méthodes de dosage du plutonium, lorsqu’on veut obtenir une plus grande sensibilité: il s’agit de l'analyse par activation et du comptage des fragments de fission. .

РАДИОХИМИЧЕСКОЕ ОПРЕДЕЛЕНИЕ ПЛУТОНИЯ ДЛЯ ЦЕЛЕЙ РАДИОЛОГИИ. В статье рассматриваются методы, которые использовались и используются в настоящее время при определении микромикроколичеств плутония в биологических образцах и об­ разцах из окружающей среды. Особое внимание уделяется анализам выделений.' Предпо­

* Work performed under Contract No. AT(45-1)-1350 between the U.S. Atomic Energy Commission and the General Electric Company.

245 246 J. M. NIELSEN and T. M. BEASLEY

лагаемая скорость выделения с мочой высчитывалась на основе предполагаемых уровней отложения плутония с тем, чтобы можно было сравнить аналитическую чувствительность различных методов. Дается описание методов полного растворения выделений, почвы, костей, тканей и растительности, причем подчеркивается необходимость избегать образования туго­ плавких плохорастворимых соединений плутония. Рассматриваются аналитические методы определения плутония в этих материалах, которые включают совместное осаждение, экстра­ гирование жидкость-жидкость, ионно-обменную хроматографию и использование изотопов плутония для конечных определений посредством анализов альфа-энергии. Посредством счетно-вычислительной статистики произведено сравнение имеющейся чувствительности в вычислении низких уровней альфа-счета с помощью ионизационной камеры, пропорциональных счетчиков, диодных счетчиков и эмульсии с ядерной меткой. Методы изотопного анализа плутония с помощью альфа-спектрометрии, методов ядерной эмульсии и жидкостных сцинтил- ляционных счетчиков также включены в этот документ. ‘ Использование неизотопных носителей в качестве источника посторонней активности и уровни содержания плутония в окружающей среде, недавно обнаруженные на земном шаре, обсуждаются в связи с контрольными образцами. Для будущих методов анализа плутония рассматриваются две возможности, где требуется повышенная чувствительность. Этими возможностями являются активационный анализ и подсчет продуктов распада.

DETERMINACIÓN RADIOQUÍMICA DEL PLUTONIO CON FINES RADIOLOGICOS. Los autores examinan los procedimientos utilizados anteriormente y en la actualidad para la determinación de cantidades de plutonio del orden de los micro-microgramos en muestras biológicas y del medio ambiente. Estudian especialmente el análisis de las excreta. Han calculado los índices de excreción urinaria probables adoptando valores arbitrarios para el depósito de plutonio, de una manera que facilita la comparación de las sensibilidades analíticas de los distintos procedimientos. Los autores describen procedimiebtos para disolver completamente muestras de excreta, suelos, huesos, tejidos y plantas, insistiendo en la.necesidad de evitar la formación de compuestos refractarios de plutonio, difíciles de disolver. Exponen métodos analíticos para efectuar el análisis del plutonio en esos materiales e incluyen la coprecipitación, la extracción liquido-liquido, la cromatografía de intercambió iónico y el empleo de los isótopos del plutonio para las determinaciones del rendimiento por análisis de la energía alfa. Recurriendo a razonamientos estadísticos, comparan las sensibilidades que es posible alcanzar en el recuento de partículas alfa de baja energía utilizando cámaras de ionización, contadores proporcionales, contadores de diodo y emulsiones nucleares. También mencionan el análisis isotópico del plutonio mediante la espectrometría alfa, las técnicas de emulsión nuclear y el recuento por centelleo líquido. . • Los autores tratan asimismo del empleo de los portadores no isotópicos como fuentes de actividad extraña y señalan los valores de la actividad del plutonio en el medio ambiente registrados recientemente en diferentes partes del mundo, relacionándolos con el empleo de muestras «en blanco». Indican dos métodos que podrían utilizarse para el análisis del plutonio cuando se requiera una sensibilidad mayor. Se trata del análisis por activación y del recuento de los fragmentos de fisión.

INTRODUCTION

The discovery of plutonium in 1940 by SEABORG, McMILLAN, KENNEDY and WAHL [1] and its subsequent production for military and peaceful uses, led to the development of chemical procedures capable of isolating and puri­ fying, submicrogram quantities of the element from a variety of matrices because of the serious radiation hazard it presents if absorbed into the body. Two methods are widely used to estimate.plutonium deposition in human beings. LANGHAM [2], has related body burden and urinary excretion of plutonium and HEALY [3 J has related lung burden and urinary excretion of plutonium. For a given detection limit for plutonium, thèse equations can be used to calculate the time during which the excretion rate will provide sufficient plutonium to exceed the detection limit by the chosen procedure. In an exposure where Langham1 s equation applies and for an annual urina­ lysis schedule, one may calculate that a procedure with a sensitivity of 0.05 dpm would allow detection of an exposure of 2.2% MPBB (maximum PLUTONIUM FOR RADIOLOGICAL PURPOSES 247

permissible body burden) even if it occurred one year previous to the ana­ lysis. If- 1%-of the MPBB must be detected, using an annual urinalysis schedule, the procedure must have a sensitivity of 0.023 dpm. If it is de­ sired to detect 1% of the plutonium MPBB by urinalysis, using a procedure with a sensitivity of 0.05 dpm, the urinalysis would then have to be made every 125'd to assure detection of cases where 1% of the body burden has been accumulated. If it is only desired to be able to detect 10% of theJMPBB, plutonium would be detectable for about 8 yr for cases in which this level of deposition had occurred. Healy1 s treatment is concerned with incidents in which insoluble plu­ tonium enters the body via the lung where it is slowly metabolized into the blood stream and subsequently eliminated in the excreta. Actual exposures are usually of a mixture of soluble and insoluble materials or of a material which is only relatively insoluble. Togèther with biological variables this factor causes excretion curve prediction to be very difficult. Some indi­ cation of this variability can be seen from data reported by SWANBERG [4]. Healy1 s treatment does, however, serve to illustrate the possibility that extremely low excretion rates may be associated with significant body burdens. . . . . Although the equation is based on an acute exposure to the lung, the clearance rate is so slow that the amount of plutonium which enters the blood stream daily decreases only slowly and the excretion rate initially increases as it would in a chronic exposure. The excretion rate passes through a maximum and decreases as the lung plutonium is depleted by dissolution. One percent MPBB in the lung with a clearance rate of 0.003 day"1 , a clear­ ance rate observed following several incidents [3] cannot be detected before 60 d or after 250 d using a bioassay procedure with a detection limit of 0.05 dpm. If the lung burden is 10% of the MPBB the quantity excreted can be determined from the time of the exposure until about 5 yr subsequent to the inhalation. Excretion rates for cases of chronic exposure are not as simply calculated as for these acute cases,., but SCHWENDIMAN et al. [5] have shown that, in general, a urinalysis procedure with a sensitivity of 0.05 dpm is adequate fo r chronic exposures. The requirements of urine analysis for plutonium have led to extremely sensitive procedures which are also applicable to other biological material, although for general health physics purposes the sensitivities required for the determination of plutonium in faeces, tissue, bone, and vegetation, are not as great as for urinalysis. However, because of the lack of knowledge of the metabolic behaviour of plutonium in humans and environmental ma­ terials, most contamination incidents, however minor, are studied to de­ termine distribution ratios and rates. Thus, for these and other research studies, very sensitive procedures are useful. This paper reviews and to some extent compares the procedures that have been used for the deter­ mination of micro-microgram quantities of plutonium in biological and en­ vironmental samples with special emphasis on excretion analysis.

DISSOLUTION OF BIOLOGICAL SAMPLES

The most crucial step .in the determination of plutonium in biological materials is the dissolution of the sample. Since-plutonium oxide can be 248 J. M. NIELSEN and T. M. BEASLEY

ve ry difficult to dissolve, or can be made so in operations .such as dry- ashing, care must be exercised in the initial dissolution and subsequent steps. Experience with urine samples and tissue samples into which plu­ tonium has come by metabolic processes indicates that the plutonium is in a readily-soluble form. Urine is commonly collected into a small quantity of formic acid to avoid complications from hydrolysis of the urea which would render the solution basic, perhaps resulting in loss of plutonium by adsorption on the container walls. Lung tissue, faeces, and excised tissue from wound sites is likely to contain insoluble plutonium. ■ One cannot generally be assured of complete dissolution of the plutonium in these latter samples without using hydrofluoric acid as a complexing agent. Removal of fluoride ion may be necessary later for the proper operation of subsequent chemical steps. In some cases tracer plutonium, such as plutonium-236 or plutonium- 238, is added at the dissolution step so that yields may be determined. As is usual in using tracers one must be certain that the tracer isotope is'in the same chemical form as the isotope being analysed before any separations are effected. This usually requires complete dissolution and sometimes oxidation and reduction steps. Yield determinations must be done by energy analysis since ordinary alpha counting and nuclear track counting do not differentiate between alpha particles of the tracer and the material being determined. . . Dissolution of biological samples is generally accomplished by wet- asliing, or a combination of dry and wet ashing [6, 7, 8] with resultant oxi^ dation of material and some of the inorganic materials. In wet-ashing, strong oxidizing agents, such as perchloric acid and nitric acid are used, either individually or in combination, or. sulphuric acid and hydrogen peroxide. Under wet-ashing conditions, lower temperatures are maintained and there is less likelihood of forming insoluble plutonium com­ pounds than with dry-ashing, where oxygen of the air serves as the oxidant. Urine and blood samples can be dissolved by treatment with concentrated nitric acid at elevated temperatures [2, 5, 9, 10]. If the plutonium is presënt as the metal, treatment with nictric acid may render it passive so that small quantities of hydrofluoric acid are required to effect its dissolution [11]. Plutonium oxides can generally be dissolved with concentrated nitric, per­ chloric and sulphuric acids, but hydrofluoric acid may also be required if the oxide is difficultly soluble. The tendency of plutonium in solution to hydro­ lyse is avoided in the analysis of biological materials since high acid con­ centrations are used. ' The oxidation of faeces, tissue, and vegetation requires repeated treat­ ment with the oxidizing agents mentioned due to the large quantities of or­ ganic material present. Multiple wet-ashings, using concentrated nitric acid and concentrated nitric acid-hydrogen peroxide mixtures have been used to oxidize faeces [9, 12, 13, 14],tissue [9, 12, 13, 14] and vegetation [15] prior to chemical analysis. Where difficultly soluble residues remain, fol­ lowing wet- and dry-ashing treatment, fusion can be used to effect solution [16]. In this technique the residue from repeated wet-ashings or the ash resulting from a dry-ashing, are fused with gram quantities of potassium or sodium carbonate in a muffle furnace until the melt is clear. The melt is then cooled to room temperature and dissolved in water dilute acid. PLUTONIUM FOR RADIOLOGICAL PURPOSES 24 9

Bone requires more drastic conditions for dissolution. Dry-ashing for long periods of time at elevated temperatures'(500-600°C) yields an ash that is soluble in hydrochloric and nitric acids [8, 16], but extrem e temperatures (700-900°C) must be avoided to prevent the possible formation of refractory plutonium compounds which are difficutly soluble. '

ANALYTICAL PROCEDURES ■ '

The procedures developed for the determination of small quantities of plutonium generally include the following steps: (a) concentration by preci­ pitation with non-isotopic carrier or by solvent extraction directly from the urine or after wet- and/or dry-ashing; (b) purification by precipitation, liquid extraction or ion exchange; and (c) determination of the plutonium content by alpha counting techniques. .

URINE

The early urinalysis procedures were based on the discovery that plu­ tonium could be quantitatively carried in the tri- and tetravalent states on bismuth phosphate and lanthanum fluoride precipitates; and these two steps were used by RUSSELL [17]. After ashing the urine with nitric acid, the plutonium is co-precipitated with bismuth phosphate. This precipitate is dissolved in hydrochloric acid, any hexavalent plutonium is reduced with sulphurous acid, and the plutonium is co-precipitated with lanthanum fluo­ ride. After transfer of the precipitate to a counting dish, the alpha activity was measured with an alpha proportional counter. Yields of 81-90% were obtained. . • ‘ ■ In a modification of this procedure [18], direct precipitation of the plu­ tonium from urine is employed using bismuth phosphate to eliminate.the need for evaporation and subsequent ashing of the residue from relatively large volumes (1000-2000 ml) of liquid. The urine specimen is made 0.2 N in n itric acid and 0.1 M in phosphoric acid, and bismuth n itrate solution is added to precipitate bismuth phosphate. The precipitate is digested over­ night, then separated and dissolved in a small volume of concentrated nitric acid. A small volume of 6% sulphurous acid is added to reduce any hexa­ valent plutonium and a second bismuth phosphate precipitation is made. Fol­ lowing dissolution of this precipitate a lanthanum fluoride precipitation is performed, and this precipitate subsequently evaporated to dryness with perchloric acid to destroy any residual organic material. The residue is dissolved in hydrochloric acid, hydroxylamine-hydrochloride is added to reduce any hexavalent plutonium that might have been formed, and a final lanthanum fluoride precipitation is performed to isolate the plutonium for alpha counting. Yields of 87-94% were reported. In recent studies, CAMPBELL and MOSS [19] found that bismuth phos­ phate co-precipitation did not give recoveries of metabolized plutonium above 50% even after predigestion with 0.5 to 2 N nitric acid for various lengths of time. • . ' , 250 J. M. NIELSEN and T. M. BEASLEY

KOSHLAND et al.[10] used calcium oxalate to co-precipitate plutonium directly from urine. Following oxidation of the oxalate, plutonium is co­ precipitated with lanthanum hydroxide to rem ove calcium. A final lanthanum fluoride precipitation is made, then the precipitate is mounted for counting. This procedure gave a yield of 92%. COOK and JONES [20] co-precipitated metabolized plutonium on calcium and-magnesium ammonium phosphates from an eight-hour sample of fresh urine. SHEEHAN et al.[21 ] found that use of this precipitation directly on urine containing metabolized plutonium gave recoveries which ranged from 85% .to as low as 6% even though consistent recoveries above 90% were ob­ tained from spiked urine samples by this procedure. By making the urine 1.5 molar in nitric acid and heating at 85-90°C for three hours, consistent recoveries above 85% were obtained. This experience led them to conclude that urinalysis procedures developed on the basis of pooled, spiked, or otherwise synthetic samples must be regarded as tentative until they have been applied successfully to samples containing metabolized plutonium. LANGHAM [22] used chloroform extraction of the plutonium cupferride to make the initial isolation of plutonium. The urine is wet-ashed with con­ centrated nitric acid until the remaining salts are white. After ignition over a burner, the salts are dissolved and the plutonium extracted from a weak hydrochloric acid solution containing ferric iron with five 3-ml portions of a 6% cupferron-chloroform solution. The combined organic fractions are evaporated to dryness and treated with concentrated nitric-perchloric acid solution at 190-200°C for about one hour to remove any organic material present. The resulting solution is diluted with a small volume of water and any hexavalent plutonium present is reduced with-hydroxylamine-hydrochloride. A final lanthanum fluoride precipitation serves to isolate the plutonium, the precipitate being transferred to a stainless steel plate, ignited, and the residue counted in an alpha proportional counter. SMALES et al. [23], also used cupferron extraction for the method he developed which, with modi­ fications, has been widely used in Great Britain [24, 25]. DALTON [26] adopted a procedure which involves an initial precipitation on calcium and magnesium ammonium phosphate, a separation by cupferron extraction, electrodeposition of the hexavalent plutonium from a basic solution, and final assay by nuclear track counting. A procedure was developed at HANFORD [5, 27] to take advantage of the greater se'nsitivities possible through the use. of nuclear emulsion track counting and the great selectivity of the chelating agent 2-thenoyl-trifluoro- acetone (TTA). The sample is wet-ashed with nitric acid and the residue heated in a muffle furnace at 575°C for one hour to insure complete oxidation of thé organic matter present. The solids from this step are dissolved in dilute nitric acid, treated with hydroxylamine-hydrochloride to reduce any hexavalent plutonium to the lower oxidation states and two'lanthanum fluoride precipitations made; The latter precipitate is separated by centrifugation, dissolved in 2 M aluminium nitrate to complex the fluoride present and sodium nitrite solution added to adjust plutonium to the tetravalent state. Plutonium is separated by solvent extraction using TTA in benzene; The plutonium is removed from the organic phase by shaking with 8 N hydro­ chloric acid. This acid solution is reduced iri volume by boiling, made basic with potassium hydroxide and the plutonium oxidized to the hexavalent state PLUTONIUM FOR RADIOLOGICAL PURPOSES 251 with sodium hypochlorite. Plutonium is electrodeposited from the potassium hydroxide solution onto a stainless steel disc. r The stainless steel disc can be counted in an alpha proportional counter or, in the case of low levels of plutonium, the disc is exposed to nuclear track film for 10 000 min (~1 week) The film is then processed and the number of tracks produced by the alpha particlés are counted with a microscope using 430X magnification. Yields of 90% are routinely obtained with a detection lim it of 0.05 dpm/1.5 1 of urine. An Expert. Committee on Methods of Radiochemical Analysis of the Joint World Health Organization/Food and Agriculture Organization recom­ mended the Hanford procedure as being the most practical for this type of work [28]. Various modifications of the Hanford procedure are in use and have been published. SANDERS [29] employed bismuth phosphate as well as lan­ thanum fluoride precipitations from acidified urine as a method of concen­ trating the plutonium. These precipitation Bteps and a perchloric acid fuming of the lanthanum fluoride precipitate serve to remove organic ma­ terial which in the original procedure is accomplished by wet-ashing and muffling. Another adaptation of the procedure is similar to Sanders' but includes an ammonium hydroxide precipitation [30]. Recoveries of 85 ± 5% are reported with a detection limit of 0.05 dpm/1.5 1 of urine. A new method involving solvent extraction was reported by BRUENGER et al. [31], which has been successfully applied to metabolized plutonium in urine and bone of dogs, but which was not studied at the low levels normally encountered in routine bioassay. After digestion in 2 N sulphuric acid the plutonium is removed by two extractions with a mixture of tert-alkyl primary amines (Primene JM-T) in xylene. The plutonium is back extracted with 8 M' HCl, dried and flam ed at 500°C, then dissolved in n itric acid and plated for counting. About 96 ±3% recovery of spiked samples was obtained. This procedure shows promise since it is simple, has a low degree of inter­ ference-of impurities (especially phosphate), and should have a high specificity. ’ Co-crystallization with potassium rhodizonate in the presence of alcohol has been used by WEISS and SHIPMAN [32] to concentrate plutonium directly from urine. Further purification .of the plutonium is. effected by co­ precipitation on lanthanum fluoride, followed by adsorption on an anion ex­ change column from concentrated hydrochloric acid. The plutonium is eluted with a hydrochloric-hydrofluoric acid mixture, co-crystallized with sodium chloride^ and after fuming with perchloric acid and sulphuric acid, the plu­ tonium is electrodeposited onto a tantalum disc from a hydrochloric acid- ammonium chloride solution. Determination of the plutonium can be ac- , complished by either alpha counting or byautoradiography. Yields of 91.3 ±5% are reported. . . / Several other ion exchange procedures have been developed for the de­ termination of plutonium in urine. SANDERS and LEIDT [33] developed a procedure which was further improved by JACOBSEN [34] in which plu­ tonium was separated from ashed urine using anion exchange chromatogra­ phy from an 8 M nitric acid solution. Plutonium is eluted with sulphurous acid, and the plutonium electrodeposited on a stainless steel disc from a weak ammonium nitrate solution. The reported yields vary from 62% at a plutonium concentration of 0.007 dpm/250 ml to 74% at a plutonium con­ centration of 0.146 dpm/250 ml. 252 J. M. NIELSEN and T. M. BEASLEY

TORIBARA et al. [16] employed anion exchange to separate plutonium from several types of biological samples. In the case of urine, repeated wet-ashing was accomplished.with concentrated nitric acid and hydrogen peroxide. After solution in 8 N hydrochloric acid, the plutonium is ad­ sorbed on a strong anion exchange resin. The plutonium is eluted with sul­ phurous acid and determined by liquid scintillation counting techniques,with yields of 95-101% reported. . CAMPBELL and MOSS [19] isolate plutonium by an alkaline earth phos­ phate co-precipitation. The precipitate is dissolved in 7.5 N nitric acid and the plutonium absorbed onto Dowex 1X2 anion exchange resin. Interfering anions are removed with 12 N hydrochloric acid and the plutonium is eluted with hydrochloric and hydroiodic acids. Alpha scintillation [35] or nuclear emulsion counting techniques are used following direct plating or electro­ deposition. A yield of 86 ±7.8% is obtained and a sensitivity comparable to that of the Hanford TTA procedure. Although this procedure requires the same time for analysing a sample as the Hanford procedure, it is said to' be simpler and require less critical control. BAINS [25] has reported results of methods development work for the determination of plutonium in urine, faeces and biological materials. As a result of these studies preferred methods were adopted. In the urine pro­ cedure, plutonium is co-precipitated with calcium and magnesium ammo­ nium phosphates and ashed. Following solution in hydrochloric acid and addition of iron carrier, the. plutonium is reduced to the tetravalent state, converted to the cupferride and extracted into chloroform. Anion exchange is used to separate plutonium from iron and other impurities and then the plutonium is' electrodeposited onto a stainless steel disc from hydrochloric acid plus ammonium oxalate. A yield of 85% was obtained. In spite of this fairly rigorous radiochemical purification, a counter employing alpha par­ ticle energy discrimination was requiredjto obtain a very low limit of de­ tection (~0.04 dpm) because of interference from natural alpha emitters which carry through the procedure. The foregoing listing of procedures does not include all which have been used and yet represents such a wide variety that it seems reasonable to sus­ pect some to be better than others and for there perhaps to be a best pro­ cedure. Several recent studies have compared and evaluated separation steps and techniques from these procedures [19, 21, 25, 26] without coming to such an agreement. It seems clear that in the field of low -level plutonium analysis the techniques are so exacting and the evaluation so difficult that it is common for the chemist to be unable to duplicate the work of others. It becomes easy to look negatively upon the work of others and often to be prem aturely confident and optim istic about his own procedure or techniques with which he is familiar. In addition, chemists place different emphasis on factors such as cost, use of certain reagents (perchloric acid, HF, ben­ zene, etc. ), simple techniques for less well-trained laboratory personnel, time for analysis, sample storage for record purposes, etc. which strongly influence choice of procedures. These factors notwithstanding, the pro­ cedures are not equivalent in certain respects which strongly affect choice. The most important of these factors are sensitivity and specificity. For high sensitivity, diode counting or nuclear track counting must be employed. This will be discussed briefly in a later section. Some of the procedures PLUTONIUM FOR RADIOLOGICAL PURPÔSES 253 do not separate the plutonium from some natural alpha em itters which inter­ fere at the very low levels, nor from some of the other actinide elements. BAENS [25] reported that some naturally-occurring alpha emitter followed the plutonium in his procedure. He found contamination with a h alf-life indi­ cating Pb212 and its daughters Bi212 and Po212 following cupferron extraction; Th228 also partly extracted. SANDERS and LEIDT [33] had indications of an intrinsic blank of 0.027 dpm/1.5 1 which they speculated may have been pu239. This has not been reported by other workers. They also noticed tracks in the nuclear emulsion twice the length of those from normal Pu239 and which they attributed to Po212 or Po214. BROOKS's procedure [24] carried an alpha emitter which appeared to be Po210. This procedure also carried some neptunium if present. This may be a weakness of several of the anion exchange procedures [36]. The specificity of the Hanford TTA procedure is such that interferences, such as reported above, have not been o b served .,

FAECES ‘ '

The levels of plutonium encountered in faeces following an inhalation exposure, are usually orders of magnitude làrger than those found in urine.. These levels can be so extremely large that faecal analysis should be per­ formed in laboratory facilities well removed from the low level urinalysis facilities to avoid serious problems of cross contamination. Following the early, rather rapid, clearance of plutonium from the lung in an acute ex­ posure, the faecal excretion data can be very useful in estimating the sys­ temic burden of plutonium [2, 3]. The methods outlined for urine can be applied to faeces using minor modifications. The principal difficulty in this analysis is to effect complete oxidation of the organic matter present. Repeated ashings with concentrated nitric acid, mixtures of nitric acid- hydrogen peroxide, nitric acid-perchloric acid or sulphuric acid-hydrogen peroxide usually effect dissolution [2, Ô, 13, 14]. D ry-ashing and bisulphate and carbonate fusions [16] may be required to effect complete oxidation and dissolution. Bains reported that a separation of iron was sometimes ne­ cessary from faecal samples if cupferron extraction was to be used [25].

BONE '

The determination of plutonium in.bone can generally be done by modified urinalysis procedures. Dry-ashing generally results in a residue which is soluble in hydrochloric and nitric acids. The relatively large amounts of calcium present in solution do not interfere in the determination of plutonium using anion exchange techniques [16] and in those procedures using precipitations with bismuth phosphate followed by lanthanum fluoride as a method of concentrating plutonium, the majority of the calcium remains in the supernatant solution [37]. ' . 254 J. M. NIELSEN and T. M. BEASLEY

SOIL '

Soil samples are usually prepared for analysis by treatment with hydro- . fluoric and perchloric acids. If insoluble residues remain, carbonate and fluoride-pyrosulphate fusions are used to insure complete solution [38]. Hydrofluoric acid as an agent to effect soil dissolution should be used with the understanding that volatilization of hexavalent plutonium from hot fluoride solutions may be encountered. Following dissolution, the procedures out­ lined previously can be used to determine the. concentration of plutoniiim. GEIGER [15] has determined plutonium in soil, vegetation, and water based on the extraction of plutonium and uranium from 4-6 N nitric acid into 50% tri-^n-butyl phosphate in n-tetradecane diluent. Average recoveries are 76 ±16% for soil, 76 ± 14% for vegetation, and 82 ± 15% for water. Dissolution techniques did not bring all of the residue into solution.

TISSUE, VEGETATION, WATER, AND AIR FILTERS .

Tissue, vegetation, and air filter samples can generally be brought into solution by the techniques described in the section on dissolution of biological samples. Water samples are generally digested the same as other samples because of the likelihood of colloidal suspensions or other solid material. The'procedures used in the analysis of urine and faeces for plutonium are equálly applicable to these samples [9, 13, 14, 16, 39]. In some cases simpler procedures are adequate for water. KOOI and HOLLSTEEN [37] have determined trace quantities of plutonium in water using a bismuth phosphate precipitation and chloroform extraction with ferric iron serving as a ca rrier for the plutonium.. Following ashing of the organic solution, the residue is dissolved in concentrated hydrochloric acid, trans­ ferred to a platinum dish and evaporated to dryness. Ferric hydroxide is precipitated by dissolving the residue in one m illiliter of water, then intro­ ducing ammonia. The resulting precipitate is dried and ignited to the oxide for counting.' Concentrations of plutonium as low as 10"10 дс/ml have been determined with 100% yield. ") . Approximate levels of plutonium contamination in air can be determined by monitoring apparatus but if high sensitivity is required radiochemical analysis is used. Since filters are generally made of organic material, and dust from the air is also collected, dissolution such as required for soil samples is often necessary. It has been estimated that 0.5 megacurie of plutonium-239 has been re­ leased to the environment as a result of nuclear weapons testing through 1958 [40]. Subsequent testing has increased this amount. Air concentra­ tions of approximately 0.2 dpm/100 m3 were recorded in Washington, D. C. [41] and Great Britain [42] during the early months of 1959. Vegetation samples [41] during the same period of time showed plutonium concentrations of 0.3-2. 4 dpm/g. Mammalian tissue samples [41], although lower in plu­ tonium concentration, showed measurable quantities rangingfrom 0.001 dpm/g for human kidney to 0.011 dpm/g for human pulmonary lymph nodes. These data serve to illustrate the fact that so-called "blank" samples can actually contain measurable quantities of plutonium. PLUTONIUM FOR RADIOLOGICAL PURPOSES 255

COUNTING AND STATISTICS

, Preparation-of the plutonium in a form suitable for counting is usually accomplished by direct plating or electrodeposition. Electrodeposition is superior to direct plating under the conditions of high salt Content since the electrodeposition procedure effects a considerable separation, from much of the remaining salts present. Electrodeposition also provides a uniform plutonium deposit which, after, flaming, is not subject to loss in handling and is capable of withstanding periods of storage for record purposes. In addition to the references already cited, LOVERIDGE [43] published a review of electrodeposition procedures in which he concluded that electrodeposition of trivalent plutonium from a slightly acid oxalate media appeared to be the most efficient. Ammonium ions must be present. . Extra care must be taken in the preparation of plates to be used for alpha energy analysis because the presence of salt deposit may degradethe spectra obtained, making interpretation difficult. This effect is especially noted when the plating area is lim ited to less than about a 10 mm2 area with the resultant areal concentration of the salt deposit. This .deposit consists largely of an oxide of iron and improved plates are obtained if an, iron ex­ traction step precedes the electrodepo.sition [25]. , In the development of a modified Hanford procedure to allow electro­ deposition of plutonium in small areas [44], it was found that the background varied considerably depending on the rare earth carrier used to make the fluoride precipitation. Of the rare earths tested, holmium and praseo­ dymium contributed least to the backgrounds while contributions from the other rare earths increased in the order dysprosium, yttrium, lanthanum and samarium. The background with samarium as a c a rrie r was about a hundredfold that with holmium or praseodymium. Ionization chambers [45], proportional counters [46], scintillation coun­ ters [47], diode counters [48], and nuclear track emulsions [26, 27] have all been used for low-level counting of plutonium. Typical backgrounds which may be achieved are 0.05 cpm for the proportional counter, 0.01 cpm for ZnS phosphor scintillation counter, 0.2 cpm for liquid scintillation counter, and 0.0015 tracks per minute for miclear track counting. Using pulse height analysis techniques, backgrounds of 0.003 cpm may.be obtained from Frisch grid ionization detectors and 0.0005 cpm from silicon junction diode detec­ tors. These latter two detectors, with their energy analysis capabilities, are able to identify the alpha-emitting nuclide from the energy of its alpha particles. This is especially advantageous where other isotopes may be present. ■ . Because of the practical restriction of reasonable counting times, only a few total counts are obtained in background and low-level sample measure­ ments. For example, using a 1000 min counting period, background values for the various types of counters listed would vary from about 1 to 200. Sta­ tistical methods must be used to evaluate the significance and probable errors involved in these counting measurements. For a total number, events of. about 30 or less Poisson statistics are used. When nuclear track counting is used, as in the Hanford procedure, with exposure times of 1.0 000 min (~ 1 week) about 1.5 alpha tracks in one-eighth of the total area which' is scanned is the average background value observed. 256 J. M. NIELSEN and T. M. BEASLEY

Any sample which results in five or more tracks has less than 1% proba­ bility of being a "blank" sample; therefore, samples giving five or more tracks have a 99% probability of being "positive". In this procedure five tracks correspond to about 0.012 dpm per sample. ■ For anàlytical purposes it is necessary to know that quantity which, if actually present, would give a positive result above the blank distribution in a given percentage of the samples analysed. This quantity is known as the "detection lim it" ànd is generally chosen to give positive values in 99% of the samples analysed. A Poisson distribution about a mean of 13 tracks would have 99% of the values above the positive sample number of five tracks. Thus, 13 tracks or 0.033 dpm is the detection limit under the conditions of this analysis. The significance and usefulness of the positive sample and detection limit concept are considered in greater detail by SCHWENDIMAN and H E A L Y [27] and H E A L Y [3]. 'With silicon junction diode detectors of 1.4-cm diam. and Frisch grid ionization chambers, backgrounds of 0.5-3 counts per day (0.0003-0.0021 cpm) can be achieved by using electronic discriminators to limit the energies of alphas counted to a 170 keV channel centred on the plutonium alpha peak. The best counting efficiencies using diode detectors, which have such limited sensitive areas, are obtained if the plutonium is plated on a small area. Efficiencies of >30% can be obtained if the diameter can be made as small as 1/8 in. The Frisch grid chambers are capable of counting larger area sources with an efficiency of about 50%. Using these values and assuming a yield of about 90% through the chemical procedure, the background level would correspond to 0.002 to 0.008 dpm, very close to the mean blank of 0.004 dpm indicated for the nuclear track method. Since one detector would be required per sample, electronic counting would probably be limited to something like an overnight (1000 min) count: This would give a total back­ ground of 1-2 counts. Using 1X10*3 cpm as the average background, a de­ tection limit of about 0.03 dpm is obtained. Thus, a detection limit equi­ valent to that obtained by nuclear track counting is realized. It should be noted that the Hanford procedure of nuclear track counting only uses about one-eighth of the area, which helps explain why diode or Frisch grid de­ tectors for 1000 min are about as sensitive as the nuclear track method for 10 000 min. The electronic counting methods require more costly equipment with a few samples per day capability at this low detector limit, while the nuclear track method is easily adaptable to larger numbers of samples. The electronic counting methods, however, have the important advantage of ener­ gy analysis which positively identifies the plutonium isotope being counted. In addition, yield determinations from added plutonium isotopic tracers can be made, and determination of the isotopic composition of the plutonium is possible. This latter capability may prove to be of importance in identifying the source of contamination for incidents for which this information is not known. • Frisch grid chamber alpha energy analysis of sample discs prepared by the standard Hanford procedure from urine containing metabolized plu­ tonium were counted with an efficiency of about 30% and a resolution of 4-5% (full width at half maximum). These samples were at the 0.1 to 6 dpm level, and were given no special treatment. Similar alpha energy analysis was reported by BAINS [25] for his procedure.. He obtained a resolution of 7%. PLUTONIUM FOR RADIOLOGICAL PURPOSES 257

By adding an iron extraction step, a resolution of 2% was obtained at plu­ tonium levels apparently within a factor of 10 of those mentioned above.'

ISOTOPIC ANALYSIS OF PLUTONIUM

The procedures discussed in-this paper have been mainly concerned with plutonium-239. Increasing production and use of other plutonium iso­ topes will require procedures to identify and measure many of them. Of those most likely to be of interest plutonium-236, -238, -239, -240, -241, and -242 are alpha emitters. Plutonium-239 can be determined by present alpha energy analysis techniques in the presence of all of these except plutonium-240, which has an alpha energy nearly the same as that for plutonium-239. Plutonium-239 : 240 ratios have been determined using a nuclear emulsion technique [49] in which the plutonium is absorbed directly into the emulsion and the plutonium-239 : 240 ratio determined by alpha and conversion electron track counting. Plutonium-238 interferes in this par­ ticular method but can be compensated for if pulse height analysis is used to determine the amount of plutonium-238 in the samples prior to exposure to the film. Plutonium-237 decays by electron capture and can be deter­ mined by gamma spectrometry. It was used'as a tracer in urinalysis studies by WEISS and SHIPMAN [32]. Plutonium-241 is a weak beta emitter which has been determined by liquid scintillation' counting following separation from aqueous media [50] and urine [51]. In the latter case, plutonium is recovered from the stainless steel discs used.in the autoradiographic de­ termination of the alpha-emitting plutonium isotopes so that plutonium-241 may be determined on samples already analysed for plutonium-239. Plu­ tonium recoveries of 85 ±9% have been obtained using spiked samples and with counting times of 24-60 h plutonium-241 can be reliably detected at the 2.2 X10"6 цс le v e l. '

. DETERMINATION OF PLUTONIUM BY ACTIVATION ANALYSIS AND FISSION FRAGMENT COUNTING

Although nuclear emulsion techniques are the most sensitive methods of analysis currently in routine use for plutonium determination, it may be desirable in some instances to measure even smaller quantities of plu­ tonium. Activation analysis and fission fragment counting offer two possi­ bilities by which this may be accomplished. McKAY [52] examined the possibility of using total fission product counting or selected isotope sepa­ ration and determination following neutron activation. If plutonium-239 is irradiated at a thermal flux of 1014n s_1cm_2for a time equivalent to 3-4 half-lives of a daughter fission product having a yield of 6%, the daughter would be present at a concentration of 5X103 dpm/dpm Pu239. If 5 dpm of this daughter is detectable, this method would theoretically provide a sensi­ tivity of 10"3 dpm Pu239; however, the feasibility of measurements at this level in the presence of expected neutron activation products has not been demonstrated. .

17 258 J. M. NIELSEN and- Т. M. BEASLEY

A fission fragment technique [53] developed for the determination of uranium in small quantities should be equally applicable to plutonium. The method is based on the fact that fission fragments leave tracks in mica which can be counted by a microscope following etching with hydrofluoric acid. The method is non-destructive and can be applied to a variety of samples. Both activation analysis and fission fragment counting require a high degree of decontamination from fissionable materials other than plutonium, and the use of facilities offering high thermal neutron fluxes, but their potential for measuring extremely small quantities of plutonium make their use attractive. ' '

' REFERENCES '

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at Hanford, Health Physics 8 (1962) 761. . . ‘ [5] SCHWENDIMAN, L. C. , HEALY, J. W. and REID, D. L. , The Application of Nuclear Track Emulsions to the Analysis of Urine for Very Low Level Plutonium, USAEC Report HW-22680, Hanford Laboratories . (November, 1951) pp. 1-47.

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Ánalytica Chimica Acta, 8 (1953) 397. - [7] HAMLIN, A. G ., Destruction of Organic Matter by Acid Ashing with Particular Reference to the Use of Perchloric Acid in the Oxidation of Textile Materials, J. Textile Industry 40 (1940)343. .

[ 8 ] GORSUCH, T. T. , Radiochemical Investigations on the Recovery for Analysis of Trace Elements in Organic and Biological Materials, Analyst 84(1959) 135. [9] RUSSELL, E. R. and NICKSON, J.J. , Distribution and Excretion of Plutonium, USAEC Report MUC-ERR- 209, University of Chicago (June 1946) pp. 1-21. [10] KOSHLAND, М. E. , BROWN, L. M. , COOK, M.J. and KOSHLAHD, D. E. , frocedure for The Determination of Plutonium in Urine, USAEC Report Mon N -92, Oak Ridge National Laboratory (May 1946) pp. 1-28. [11] KOLTHOFF, I. M. and ELVIG, P. J. (Ed.), Treatise on Analytical Chemistry 9, Interscience, New York (1962) 237. . • • • - [1 2 ] LANGHAM, W. H. , LAWRENCE, • J. N. P. , MCCLELLAND, J. and HEMPELMANN, L. H. , Th e Los Alam os

. Scientific Laboratory's Experience with Plutonium in Man, Health Physics 8 (1962) 753. , [13] STOVER, B.J. , ATHERTON, D. R. , BRUENGER, F. W. and BUSTER, D. S. , Further Studies of the Meta­

bolism of Plutonium-239 in Adult Beagles, Health Physics 8 (1962) 589. [14] BAIR, W. J. , WILLARD, О. H. , HERRING, J. P. and GEORGE, L. A ., Retention, Translocation and Ex­

cretion of Inhaled Pu2390 2, Health Physics 8_(1962) 639. [15] GEIGER, E. L. , Radioassay of Uranium and Plutonium in Vegetation, Soil and Water, Health Physics 1^ (1959) 405. ■ ■ . [16] TORIBARA, Т .К ., PREDMORE, Carol and HARGRAVE, Paul A. , The Separation and Determination of Plutonium in Diverse Biological Samples, Talanta 1_0 (1962) 209. ' [17] RUSSELL, E. R ., Procedures for the Determination of Plutonium in Human Urine, USAEC Report MUC- ERR-156, University of Chicago Metallurgical Laboratory (Oct. 1945) pp. 1-11. [18] FARABEE, L. B. , Procedure for the Determination.of Plutonium in Urine, USAEC Report Mon H -218, Oak Ridge National Laboratory (April 1947) pp. 1-17. [19] CAMPBELL, E. E. and MOSS, W. D. , Determination of Plutonium in Urine by Anion Exchange, Proc. 9th Ann. Bioassay and Analyt. Chemistry Mtg, San Diego, Calif., (Oct. 10-11 1963), to be published.

[2 0 ] COOK, G. P. , JONES, O. , 190 A M /V -68 (1 9 5 2 ). v PLUTONIUM FOR RADIOLOGICAL PURPOSES 259

[21] SHEEHAN, W .E. WOOD, W. R, , Jr. , and KIRBY,. H. W ., Urinalysis for Metabolized Plutonium, Proc. 9th Ann. Bioassay and Analyt. Chem. Mtg, San Diego, Calif. (Oct. 11-12, 1963),to be.published. [2 2 ] LANGHAM, W. H ., Determination of Plutonium in Human Urine, USAEC Report No. MDOC-1555, Los Alamos Scientific Laboratory (Aug. 1947) pp. 1-5. [23] SMALES, A. A ., AIREY, L .', WALTON, G. N. and BROOKS, R. O. R. , The Determination of Plutonium in Urine,- Report AERE-C/R-533, Atomic Energy Research Establishment (May 1950) pp. 1-13. [24] BROOKS, R. O. R. , Collected Laboratory Procedures for the Determination of Radioelements In Urine, Report AERE-AM-60 (1960). [25] BAINS, M. E. D. , The Determination of Plutonium Alpha Activity in Urine, Feces and Biological Ma­ terials, Report AEEW-R-292 (1963). ■ '• - [26] DALTON, J. C. , The Determination of Plutonium in Urine By Direct Phosphate Precipitation and Auto­

radiography, PG Report 284 (W) (1962). 4 [27] SCHWENDIMAN, L. C. and HEALY, J. W. , Nuclear-Track Téchnique for LowLevel Plutonium in Urine,

Nucleonics _16 6 (1958) 78. ■ [28] WORLD HEALTH ORGANIZATION, Technical Report Series, No. 173, Methods of Radiochemical Analysis, Report of a Joint WHO/FAO Expert Committee, Geneva (1959). [29] SANDERS, S. M.Jr. , Determination of Plutonium in Urine, USAEC Report DP-146, Savannah River La­ boratory (March 1956) pp. 1-12. • ' [30] MCCLELLAND, J. (Ed. ),Analytical Procedures for the Industrial Hygiene Group, USAEC Report LA-1858, Los Alamos Scientific Laboratory (Aug. 1955) pp. 155-171. [31] BRUENGER, F.W ,, STOVER, B.J. and ATHERTON, D. R ., Determination of Plutonium in Biological Material by Solvent Extraction with Primary Amines, Anal. Chem. 35 (1963) 1671. [32] WEISS, H. V. and SHIPMAN, W. H ., Radiochemical Determination of Plutonium in Urine, Anal. Chem. 33(1961) 37. [33] SANDERS, S. M ., Jr. and LEIDT, Sarah C ., A New Procedure for Plutonium Urinalysis, Health Physics §_ (1961) 189. [34] JACOBSEN, W. R ., Initial Two-Year Experience with the Savannah River Plutonium Analysis Procedure, USAEC Report ANL-6637 (1961) pp. 32“42. [35] HALLDEN, Naomi A. and HARLEY, J. H. , An Improved Alpha-Counting Technique, Anal. Chem. 32 (1960) 1861-1863. [36] WISH, L. and ROWELL, М ., Sequential Analysis of Tracer Amounts of Neptunium, Uranium and Plu­ tonium in Fission Product Mixtures by Anion Exchange, USNRDL Report, TR-117, Naval Radiological

Defense Laboratory (Oct. 1956) pp. 1 - 3 4 . - [37] KOOI, J. and HOLLSTEIN, U. , An Improved Method for the Determination of Trace Quantities of Plu­ tonium in Aqueous-Media - II, Health Physics 8^(1962) 49. [38] GRAVES, A. W. , Silicate Analysis, 2nd ed. , George Allen and Unwin, Ltd. , London (1951) p. 49. [39] SCHUBERT, J. , MYERS, L. S. , Jr. and JACKSON, J. A. , The Analytical Procedures of the Bioassay Group at the Argonne National Laboratory, USAEC Report ANL-4509, Argonne National Laboratory (March 1951) pp. 1-23. [40] UNITED NATIONS SCIENTIFIC COMMITTEE ON THE EFFECTS OF ATOMIC RADIATION, Status Report of the Food Chain, Document A/AC 821R-80. [41] KREY, P. W. , BOGEN, D. and FRENCH, E. , Plutonium in Man and His Environment, Nature 195 (1962) 26 3 . . [4 2 ] PIERSON, D. H. , CROOKS, R, N. and FISHER, E. M. R ., The R adioactivity of the Atmosphere Near Ground Level Due to Distant Nuclear Test Explosions, Report AERE-M620, Atomic Energy Research Establishment (May 1960) pp. 1-8, [43] LOVERIDGE, B. A ., Quantitative Electrodeposition of Plutonium for Alpha and Beta Assay, Report AERE- R3266 (1960). . [44] PERKINS, R.W. , Private Communication, Hanford Laboratories (June,1963). [ 4 5 ] JA F FE Y , A. H. , KOHM AN, Т . К . and CRAWFORD, J. A ., A (Manual on The Measurement of Radio­ activity, Report CC-1602 (M -626) (March 1944). [46] MCCLELLAND, J. (Ed. ),Analytical Procedures for the Industrial Hygiene Group, USAEC Report LA-1858, Los Alamos Scientific Laboratory (Aug. 1955) pp. 121-124. * [47] SMALES, A. A. , 'AIREY, L. , WOODWARD, J. and MAPPER, D. , Notes on the Use of The Scintillation Alpha Counter, AERE C/R 306 (Feb. 1949). 2 60 J. M. NIELSEN and T. M BEASLEY

[48] BLANKENSHIP, J. L. and BORKOWSKI, C. J. , Silicon Surface-Barrier Nuclear Particle Detectors, IRE , Transactions NS-7, No. 2 (June 1960) pp. 190-195,

[49] SLOTH, E. N. and STUDIER, М. H ., Nuclear Emulsion Technique for Determination of The Pu^/Pu 239 Ratio, Anal. Chem. 3£(1958) 1751. [50] HORROCKS, D. L. and STUDIER, М. H ., Low-Level Plutonium-241 Analysis by Liquid Scintillation Techniques, Anal. Chem. 30 (1958) 1757. .

[51] LUDWICK, J. D. , The Analysis of Plutonium-241 in Urine, Health Physics 6 (1961) 63, [52] McKAY, H. A. C ., Detection of Plutonium by Irradiation in the Pile, Report AERE-C-M-23, pp. 1-3, Atomic Energy Research Establishment (June 1949). [53] PRICE, P. B. and WALKER, R. M. , A Simple Method of Measuring Low Uranium Concentrations in Natural Crystals, Applied Physics Letter 2 (Jan. 15, 1963) 23-5.

DISCUSSION

F.E. BUTLER: I would like to ask Dr. Nielson about the low back­ ground of two-tenths of a count per minute he reported for liquid scintillation counting. Was this reported in his reference [47] or has he any personal experience of it? ■ J. M. NIELSON: This value represents the best low-level liquid scintil­ lation background for alpha counting obtained at the Argonne National Labora­ tory and at the Hanford Laboratories. A PROCEDURE FOR THE DETERMINATION OF ALPHA-EMITTING PLUTONIUM IN URINE USING A SOLID-STATE COUNTER

F.J. SANDALLS AND A. MORGAN . HEALTH PHYSICS AND MEDICAL DIVISION, ATOMIC ENERGY RESEARCH ESTABLISHMENT, HARWELL, ENGLAND

Abstract —■ Résumé — Аннотация — Resumen

A PROCEDURE FOR THE DETERMINATION OF ALPHA-EMITTING PLUTONIUM IN URINE USING A SOLID-STATE COUNTER. A method for the routine determination of alpha-emitting plutonium in urine is described. In evolving this procedure various techniques for concentrating, purifying and electrodepositing plutonium were compared, and these investigations are summarized. In the procedure finally adopted, urine is wet-ashed and the residue dissolved in hydrochloric acid. Plutonium (IV) is co-precipitated with iron cupferride from this solution and extracted into chloroform. After evaporation of the chloroform, the residue is oxidized, dissolved in hydrochloric acid and the iron extracted into di-isopropyl ether. The aqueous phase containing the plutonium is evaporated and dissolved in an acid ammonium sulphate solution from which the plutonium is electrodeposited onto stainless steel. Quantitative recoveries are obtained over the electro­ deposition stage and on overall recovery of 84 ± 7%. The^electrodeposited plutonium is counted with a solid-state (silicon junction diode) detector in a counter develop ed "1 for this purpose. The low inherent background of this type of counter is effectively reduced still further by counting only those alpha particles with energy in the 4.2 - 5.4 MeV range. The good resolution which can be obtained with thin electrodeposited sources of plutonium enables this narrow channel to be used with only small losses in counting efficiency. By counting over this restricted energy range, the blank activity arising from reagents and incomplete removal of alpha-emitting contaminants in urine is also reduced by a factor of two, to just over 0.01 pc/24-h sample. The limit of detection with this method is about 0.025 pc of Pu23*.

MÉTHODE DE DÉTERMINATION DU PLUTONIUM ÉMETTEUR ALPHA PRÉSENT DANS L'URINE, AU MOYEN D’UN COMPTEUR A SEMI-CONDUCTEURS. Les auteurs exposent une méthode pour la détermination courante du plutonium émetteur alpha présent dans l'urine. Pour mettre au point cette méthode, ils ont com­ paré diverse procédés de concentration, de purification et de dépôt électrolytique du plutonium; les résultats de ces recherches sont résumés dans le mémoire. La méthpde qu'ils ont finalement adoptée consiste à incinérer l'urine et à dissoudre le résidu dans de l'acide chlorhydrique. A partir de cette solution,' le plutonium (IV) est coprécipité avec du cupferrate de fer et extrait au moyen de chloroforme. Après évaporation de celui-ci, le résidu est oxydé et dissous dans l'acide chlorhydrique, puis le fer est extrait au moyen d’éther di-isopropy- lique. La phase aqueuse contenant le plutonium est évaporée et dissoute dans une solution de sulfate acide d'ammonium, d'où le plutonium est déposé par électrolyse sur de l'acier inoxydable. Au stade du dépôt . électrolytique, on recueille dans l'ensemble 84 ± 7% du plutonium contenu. Le plutonium déposé par électrolyse est mesuré au moyen d'un détecteur à semi-conducteur (diode à jonction au silicium) dans un compteur conçu à cette fin. Le faible mouvement propre de ce type de compteur est encore efficacement réduit si l'on ne compte que les particules alpha ayant des énergies comprises entre 4,2 et 5,4 MeV. La bonne résolution que l'on peut obtenir avec de minces couches de plutonium déposées par électrolyse permet d'utiliser ce canal étroit en ne faisant que de faibles pertes de rendement de comptage. En ne comptant que dans cette gamma d'énergies peu étendue, l'activité due à la présence de réactifs et d'autres agents de contamination émetteurs alpha résiduels dans l'urine, qui n’est pas entièrement déconta­ minée, se trouve également réduite de moitié et atteint un peu plus de 0 ,0 1 pc par échantillon prélevé sur les urines recueillies pendant 24 h. La limite de détection à l'aide de cette méthode est d’environ 0, 025 pc de 239Pu. .

МЕТОД ОПРЕДЕЛЕНИЯ АЛЬФА-ИЗЛУЧЕНИЯ ПЛУТОНИЯ В МОЧЕ С ПОМОЩЬЮ ПОЛУ­ ПРОВОДНИКОВОГО СЧЕТЧИКА. Описан метод практического'определения альфа-излучения

261 262 F. J. SAND ALLS and' A. MORGAN

плутония в моче. При разработке этого метода сравнивались различные методы концентри­ рования, очистки и электролитического осаждения плутония ; подводится итог этих исследований. Согласно принятому в конечном итоге методу мочу подвергают влажному озолению, осадок растворяют в соляной кислоте. Плутоний (IV) подвергают совместному осаждению с купферронатом железа из этого раствора и экстрагируют хлороформом. После выпаривания хлороформом осадок окисляют, растворяют в соляной кислоте и экстрагируют железо ди- изопропиловым эфиром. Водную фазу, содержащую плутоний, выпаривают и растворяют в кислом сульфате аммония, из которого плутоний осаждают электролитически на нержавеющей стали. Количественные восстановления на стадии электролитического осаждения дали 84 ± 7То Осажденный плутоний подсчитывали с помощью диодного (твердого с кремневым соедине­ нием) счетчика,созданного для этой цели. Низкий естественный фон счетчика этого типа был эффективно снижен путем подсчета лишь тех альфа-частиц, которь'ге имеют энергию 4.2 — 5,4 М эв. Хорошее разрешение, получаемое с помощью тонкого электроосажденного источника плутония3обеспечивает возможность очень малых потерь эффективности счета в этом узком канале. При счете вне'этой ограниченной области энергии контрольная активность, создаваемая реактивами и неполной очисткой от других альфа-излучателей в моче, также снижается на фактор порядка 2, немного выше 0,01 мкмккюри в-течение 24 часов. Предел обнаружения с помощью этого метода составляет около 0,025 мкмккюри плутония-239. .

PROCEDIMIENTO PARA DETERMINAR LA ACTIVIDAD ALFA DEL PLUTONIO EN LA ORINA CON AYUDA DE UN CONTADOR DE ESTADO SÓLIDO. La memoria describe un método para determinaciones corrientes de plutonio en la orina. También resume las investigaciones que llevaron a establecer este procedimiento y en las que se compararon diversas técnicas de concentración, purificación y depósito electrolítico del plutonio. Con arreglo al procedimiento adoptado en definitiva, la orina se calcina en húmedo y el residuo se disuelve con ácido clorhídrico. El plutonio (IV) presente en la solución asi obtenida se coprecipita con cupferrato de hierro y se extrae con cloroformo. Una vez evaporado éste,, se oxida el residuo, se disuelve en ácido clorhídrico y el hierro se extrae con éter diisopropflico. Seguidamente se evapora a sequedad la fase acuosa que contiene, al plutonio y se disuelve en una solución ácida de sulfato amónico desde la cual se separa el plutonio por depósito electrolítico sobre acero inoxidable. El rendimiento de esta última etapa es cuantita­ tivo, siendo de 84 ± 7% para.el proceso en conjunto. . El plutonio electrodepositado se recuenta con un detector de estado sólido (diode de unión de silicio) en un contador especialmente construido para este fin. La baja actividad de fondo inherente de este tipo de contadores puede disminuirse aún más contando solamente las partículas alfa de energía comprendida entre 4.2 y 5,4 MeV. El buen poder de resolución que se obtiene con capas delgadas de plutonio electrodepositado permite emplear un canal tan estrecho sin que la eficacia de recuento disminuya apreciablemente. Al efectuar el recuento en este intervalo restringido, la actividad de fondo debida a los reactivos y a la eliminación incompleta de las impurezas de emisores alfa de la orina queda reducida también a la mitad, esto es, a un valor del orden de 0,01 pcuries por muestra de 24 horas. El límite de detección mediante este

método es aproximadamente de 0, 025 pc de 239 Pu. ‘

1. INTRODUCTION

Internal exposure to plutonium-239 at or below the maximum permissi­ ble levels recommended by the ICRP [1] can only be assessed by bioassáy determinations. Methods for the in vivo determination of Pu239 have been proposed in which the associated X-rays are detected by a sodium iodide crystal or proportional counter. The abundance of these X-rays is so low, however, that such methods are only likely to be applicable to the determi­ nation of Pu239 in specific organs, such as liver or lungs, at concentrations in excess of the permissible values. If the human experimental data of LANGHAM [2] are taken as a model, it is evident that systemic plutonium is excreted only very slowly, and after two months the urinary excretion rate is only 0.01%/d of the original intake. A fter the acute intake of a maxi­ ALPHA-EMITTING PLUTONIUM IN URINE 263 mum permissible body burden (MPBB) of Pu239, this rate of excretion corre­ sponds to 4 pc/d and to detect low-level chronic exposure at an early stage it is therefore essential to be able to determine Pu239 in bioassay samples at the 0.1 pc level. Until recently the method used at AERE, Harwell, for the determination of plutonium-239 in urine was based on that originally described by SMALES et al. [3] and modified in minor respects by BROOKS [4] . In this method, urine is evaporated, ashed and the plutonium purified by со-precipitation with iron cupferride and extraction into chloroform. Finally the iron carry­ ing the plutonium is mounted on a platinum tray and counted in a conventional alpha scintillation counter with a ZnS screen. Plutonium recoveries obtained with this method are high and consistent, but significant levels of alpha- emitters arising from the reagents give a relatively high "blank" value of about 0.1 pc. The sensitivity of the method is also limited by the compara­ tively high background (~5 counts/h) of the scintillation counters used. It was evident that a reduction in counter background would result in a valuable gain in sensitivity and alternative counting techniques were con­ sidered. These included nuclear-track autoradiography, scintillation counting with ZnS screens of small area and the use of solid-state (silicon junction diode) detectors . Autoradiography was discounted as automatic track counting equipment was not available at the time and we wished to avoid the tedious manual counting of tracks. Of the other alternatives, it appeared as though the excellent resolution obtainable with the solid-state detector gave it an important advantage over.the small-screen counter. In this paper some preliminary trials with a prototype solid-state counter are described, which were undertaken to assess the advantages to be gained from the use of this type óf counter. As the preparation of thin electrodepositçd sources of plutonium is essential for the advantages of this technique to be fully realize4 concurrent work has been carried out to compare various published electro­ deposition procedures.

2. E X P E R IM E N T A L •

The analysis of plutonium in urine can be conveniently divided, into three stages; the initial concentration, the purification, and finally the m easure­ ment stage. Investigations carried out at each stage of the procedure are described below.' •

2.1. The concentration stage

Preliminary concentration of plutonium from urine may be carried out either by evaporation and wet-ashing or by co-precipitation of the plutonium on bismuth phosphate [5], potassium rhodizonate [6] or calcium and mag­ nesium animonium phosphates [7] . Co-precipitation has the obvious merit that the evaporation of large quantities of urine is avoided, but before such a method can be used, it is essential to confirm that metabolized plutonium is co-precipitated quantitatively. DALTON [7] has reported that calcium and magnesium ammonium phosphates carry down metabolized plutonium effectively, and for a time this technique was used instead of evaporation as 264 F .J . SAND ALLS and A. MORGAN the first stage in the analytical procedure. In this method, a hydrochloric acid solution containing 0.3 g of calcium phosphate is added to untreated urine and the solution made alkaline with ammonia. Plutonium is co-precipitated with the calcium and magnesium ammonium phosphates, which are removed from the urine by filtration. It was found that when urine samples from non­ exposed people were analysed by this procedure, a "blank" result of 0.15 ± 0.04 pc/24 h sample was obtained. Similar results were obtained when distilled water was analysed by the same procedure, which showed that the origin of this alpha activity was in the reagents. Similar conclusions were drawn by Dalton [7], although the "blank" levels he obtained were low er (0.016 - 0.032 pc/500 m l of urine), indicating that either the reagents he used may not have been contaminated to the .same degree, or that the electro ­ deposition stage effected additional decontamination. Examination of the reagents showed that the calcium phosphate used contained significant amounts of Ra226 and daughters and probably members of the thorium series (Bi212 and Po212) as well. Interference from these members of the thorium series has also been reported by BAINS [8] who pointed out that both lead and bismuth are known to form cupferrides which- are probably soluble in chloroform in tracer quantities. In an attempt to avoid introducing this extraneous alpha activity into the plutonium procedure, calcium phosphate was prepared by adding orthophosphoric acid to a solution of calcium nitrate.. The phosphoric acid itself turned out to be contaminated however, and FARABEE [9] has also reported the identification of radium-D and daughters in samples of phosphoric acid. Although it was found that the alpha activity of phosphoric acid solutions coùld be reduced considerably by extraction with cupferron-chloroform at à pH of 5, the co-precipitation pro­ cedure was discontinued, as the introduction of any extraneous alpha-emitter which might be carried through the procedure will increase the "blank" level and affect the sensitivity adversely. A wet oxidation stage was substituted, which in practice gives somewhat higher and more consistent recoveries of plutonium than the phosphate co-precipitation procedure.

2.2. The purification stage

To count plutonium in a solid-state counter it must first be electro­ deposited to give a thin source of small area. Before adapting the cupferron procedure already in use, two published methods already incorporating electrodeposition stages were investigated. The first of these is that de­ veloped at Hanford by SCHWENDIMAN and H EALY [10], in which plutonium is co-precipitated with lanthanum fluoride and subsequently extracted into benzene as the thenoyltrifluoroacetone (TTA) complex. The plutonium is finally electrodeposited from a potassium hydroxide-sodium hypochlorite electrolyte. One disadvantage of this procedure is that lanthanum nitrate, used in the initial со-precipitation step, contains natural alpha-emitters and must therefore be purified before use. When this procedure was tried at AERE, plutonium recoveries of 80% were obtained over the purification stages, but the overall recovery when the electrodeposition stage was added did not reach that claimed by the authors. The second method investigated was that described by SANDERS and LEIDT [11] in which an anion-exchange resin is used to separate plutonium ALPHA-EMITTING PLÜTONIUM IN URINE 265 from other material after the wet oxidation of ut-ine. Very variable re­ coveries w ere obtained with'this procedure and it was found to be extrem ely difficult to control the extremely low flow-rates specified. It should be pointed out that JACOBSEN [12] , describing the initial tw;o years' experience with this method at the Savannah R iver Plant, reports that it has been m odi­ fied in several important respects, but the revised procedure has not yet been triéd at AERE. • ' Of the three'methods investigated, the cupferron procedure gives the most satisfactory results, both in terms of recovery of plutonium and con­ venience in operation. Good decontamination from uranium is claimed by Smales et al. [3] and Bains [8] has reported that if the cupferrides are extracted into chloroform at a pH of 0.3, a reasonable degree of separation from thorium is obtained. KEMP [13] has recently studied the conditions under which Pu (III) is oxidized to Pu (IV). He found that the addition of cupferron and hydroxylamine to the solution some time before chloroform extraction, ensured the complete oxidation of plutonium to the tetravalent state, which is necessary for short extraction times. It was decided there­ fore to retain the cupferron procedure and to add an electrodeposition stage to produce a plutonium source suitable for counting with a solid-state detector.

2. 3. The electrodeposition stage

It is desirable that an electrodeposition method should give high and consistent recoveries in a reasonably short time and, in addition, if a solid- state detector is used, the deposited plutonium layer should be of small area and as thin as possible to achieve high counting efficiency and minimize the degradation of alpha particle energy ; > In the investigations described below, a number of published electrodeposition procedures were investigated using an electrodeposition cell of the type described by SCHWENDIMAN, HEALY and REID [14] unless otherwise stated. The first procedure to be examined was that of Schwendiman and Healy [10] who used a potassium hydroxide-sodium hypochlorite electrolyte and stainless-steel cathode. When plutonium was added directly to the electro­ lyte a mean recovery of 80 i 4% was obtained in 5 h. By increasing the current from 200 to 400 mA and the plating time to -5| h, Dalton [7] obtained recoveries slightly in excess of 90%.: The main drawback to this method is the long plating time required. . The second method to be tried was that devised by MITCHELL [15] for the rapid electrodeposition of plutonium. In this, plutonium is electro­ deposited from an ammonium chloride solution onto a platinum cathode (3.8 cm2 in area) using a current of 2.5 A. It was not possible toreproduce these plating conditions exactly,, but using a tantalum .cathode of area 1.1 cm2 and a'current of 500 mA, a series of experiments indicáted that deposition should be complete after 1 h: However, when a number o f trials was carried out using these conditions, variable recoveries averaging 78 .± 23% were obtained. ■ ■ • . . The next procedure to be tried was that of LOVERIDGE.[16] who claims that quantitative recoveries of.plutonium can be obtained by'electrodeposition onto a stainless-steel cathode from a slightly acid solution of .ammonium 266 F. J. SANDALLS and A. MORGAN

oxalate. The recommended time for the electrodeposition is 2\ h and using these conditions, high and consistent recoveries were obtained averaging 97 ± 7%. Although the recovery was excellent, plutonium sources prepared froni this electrolyte consisted of a black and somewhat friable deposit on the stainless steel. The most satisfactory procedure was found to be that originally described by DUPZYK and BIGGS [17] for the electrodeposition of curium. In this method, curium is deposited onto stainless steel from an ammonium sulphate electrolyte, using a current of 300 mA. Using an identical procedure for plutonium, the amount deposited as a function of electrodeposition time was studied and the results are shown in Fig. 1. It is evident that almpst quanti -

. ' - Fig. l

Recovery of plutonium from an ammonium sulphate electrolyte ■ as a function of time

tative recoveries can be obtained in 2 h. In a series of experiments in which plutonium was added directly to the electrolyte, an average recovery of .99 ± 5% was obtained and the sources prepared in this way consist of an almost invisible coherent deposit in the stainless steel. ¡From the results obtained, it appears that both the ammonium oxalate and ammonium sulphate electrodeposition procedures will give practically quantitative recoveries of plutonium, but a more satisfactory source is ob- tainèd from the latter. In both these methods, the presence of çven small amounts of iron in the electrolyte can affect the plutonium recovery ad­ versely, and if either of these methods are to be added to the cupferron pro­ cedure, it is essential to remove the iron from the oxidized residue re­ maining after evaporation of the chloroform. It was found that if this residue is dissolved in 8 N hydrochloric acid, two extractions with di-isopropyl ether will remove practically all the iron from the acid solution. Bains [8] reports that no plutonium is extracted under these conditions, but that a trace of iron is. left in the acid solution, which cannot be removed by additional extractions ALPHA-EMITTING , PLUTONIUM IN URINE 267 with the ether. The small amount remaining does not, however, appear to affect the recovery of plutonium from the'electrodeposition stage. After removal of the iron, the1 hydrochloric-acid solution is evaporated to dryness and at this stage the residue usually contains carbonaceous material which may be oxidized by heating with i ml of concentrated nitric acid and evapo­ rating to dryness. Concentrated sulphuric acid (0.3 ml) is added to the residue, heated until dense white fumes are evolved, and the solution transferred quantitatively to an electrodeposition cell with 5 ml of water and 2 drops of methyl red indicator. The solution is neutralized by the addition of ammonia and the red colour of the indicator quickly restored by the drop-wise ad­ dition of 1.5 M sulphuric acid. After adjusting the pH in this way, a current of 300 mA is passed through the cell for 2 h*. The recoveries obtained in a series of trials with spiked urine samples are given in Table I and averaged 84 ± 7%. Full details of the complete procedure have been reported by SAND ALLS and MORGAN [18] .

TA BLE I

RECOVERY OF PLUTONIUM-239 FROM SPIKED URINE SAMPLES

R eco v ery Sample number СЙ

. 1 8 8 . 8

. 2 . 8 6 . 6 3 7 2 . 6 4 8 5 . 9 5 7 1 . 2

6 8 4 . 2 7 8 8 . 8 ■'

8 . v 9 0 .8 9 9 3 . 2

10 7 8 . 0

11 ■ 8 4 . 8

12 8 4 . 2

’ Mean and SD - 8 4 ± 4% . ■'

3. EXPERIMENTS WITH A PROTOTYPE SOLID-STATE COUNTER

The main advantages to be obtained from the use of solid-state detectors are their low intrinsic background and excellent resolution characteristics. By incorporating them in a counter with suitably set discriminators, they may be used to count only those alpha particles which fall within a pre­ determined energy band or channel. The width to which this channel may be reduced, without significant loss in detection efficiency, depends upon the degradation of alpha particle energy occurring within the source and in the

^ Recent work at the Atomic Weapons Research Establishment (AWRE) Aldermaston, has shown that more consistent recoveries are obtained .by increasing the electodeposition tim e to 3 h. 2 6 8 F. J. SANDALLS and A. MORGAN ' air-gap between the source and detector. The latter effect can be made almost negligible by positioning the source very close to the face of the detector. The electrodeposited source itself will have a finite thickness however, particularly if prepared from impure solutions, and this is where most of the degradation occurs. The thinner the source, the narrower the plutonium channel can be made and the narrower the channel, the less extra­ neous alpha activity will be counted, irrespective of whether it is part of the intrinsic counter background, alpha activity arising from the reagents, or from incomplete decontamination from other alpha emitters occurring in urine. To determine how narrow this plutonium channel can be made, plu­ tonium sources were prepared by adding Pu239 to the ammonium sulphate electrolyte before electrodeposition onto stainless steel and also by analysing urine to which Pu239 had been added. The alpha spectra of sources prepared by each of these methods were measured with a pulse-height analyser oper­ ating in conjunction with a solid-state detector. The spectrum obtained when Pu239 is added directly to the electrolyte is shown in Fig. 2. The deposit is

F ig . 2

Alpha spectrum of an electrodeposited source prepared by adding

Pu239 directly to the electrolyte invisible and the resolution obtained (about 2%) is very close to the optimum which can be achieved in air. The spectrum obtained from the source pre­ pared from spiked urine is shown in Fig. 3. The resolution in this case is about 6% and the deposit is visible as a very thin grey-black film on the stain­ less steel. The thickness of the film has been found, by weighing, to be about 400-500 /ug/cm2 and the degradation of the Pu239 spectrum is that which would be expected from a source of this thickness.. This spectrum shows that a channel width of 1.2 MeV can be used to count.electrodeposited Pu239 sources prepared from urine, with little loss in counting efficiency. Under optimum conditions, with very thin sources, the- channel could be reduced to half this width, which would give a corresponding reduction in background. ALPHA-EMITTING PLUTONIUM IN URINE 269

F ig - 3

Alpha spectrum of an electrodeposited source prepared from urine

to which Pu 239 had been added .

3. 1. Effect of discrimination on counter background

A prototype counter was designed and built in the Electronics Division at AERE. This is shown in Fig. 4 and has been described in detail by MUSTON 119J . Three variable level discriminators are incorporated in the counter, so that each of the three registers records only those alpha parti­ cles with energy exceeding a predetermined value. The first register records all alpha particles with energy exceeding 2 MeV; the second, those with energy exceeding 4.2 MeV and the third, those exceeding 5.4 MeV. As shown in Fig. 3, practically all the alpha particles from electrodeposited sources of Pu239 prepared from urine will be counted by the second register and very few by the third. , .

The results of a series of background measurements made with a blank stainless-steel disc under the detector are shown in Table II, together with the mean total counting rates and counting rates in the plutonium channel. The background in the plutonium channel is one-third of the total background, so that by counting in this region only, the background counting rate is re­ duced by a factor o f three to 0.14 counts/h.

To determine the efficiency of the counter operating under these con­ ditions, a series' of urine samples, to which a known amount of Pu239 had been added, were analysed. The efficiency of the counter, calculated from the total counting rate (>2 MeV) is 35%, while the efficiency based upon the counting rate in the plutonium channel (4.2 - 5.4 MeV) is 28%, representing a loss in efficiency of only 7%. The reduction in background more than compensates for this loss in efficiency. 2 70 F. J. SAND ALLS and A. MORGAN

F ig . 4

Prototype solid-state counter showing detector and source holder

3.2. Measurements on urine from non-exposed people

When urine samples from nominally unexposed people are analysed for Pu239 by the procedure described in this paper, the final electrodeposited source has a small, but significant, alpha activity. To establish the level of this "blank" activity and the extent to which it is reduced by the use of discrimination, twelve 1.5 1 urine samples from unexposed personnel were analysed. Six of these were obtained by taking 1.5 1 aliquots (samples 1-6) from pooled urine sent for tritium analysis and six were individual 1.51 samples (7-12) obtained from members of the AERE Bioassay Section. The results of these determinations are given in Table III. The counter background was deducted before calculating the activity. There is no significant difference TABLE II

COUNTER BACKGROUND IN THE >2 MeV AND 4.2 - 5.4 MeV RANGES LH-MTIG LTNU I UIE 1 7 2 URINE IN PLUTONIUM ALPHA-EMITTING

. Counts recorded Background counting rate C o u n tin g period Plu ton iu m > 2 M eV > 4 . 2 MeV > 5 . 4 M eV 4. 2-5.4 MeV T o ta l (h) ch a n n e l (a) (b) ( c ) . .(b - C) (co u n ts/h ) (co u n ts/h )

16 7 . 4 1 '■ 3 0 . 4 4 0 . 1 9 ■ .1 6 2 . 53 2 8 4 ■ 2 4 0 . 3 3 0 . 1 5

1 6 6 2 ' 2 • 0 0 . 3 8 : - • 0 . 0 0 ■ ■

1 1 2 37 23 8 1 5 ' 0 . 3 3 0 . 1 3

96 57 29 . 1 8 ‘ 11 0 . 5 7 0 . 1 1 1 4 4 . 73 3 0 . 4 • 2 6 0 . 51 0 . 1 8

9 3 38 19 11 8 ( 0 . 4 1 0 . 0 9

_ M ean 0 . 4 2 0 . 1 4 272 F. J. SANDALLS and A. MORGAN

TA BLE III

BLANK DETERMINATIONS ON URINE SAMPLES FROM NON-EXPOSED PEOPLE

Result Result S am p le S a m p le • (р с Д . 5 1) ( p c / 1 . 5 1)

1 0 . 0 1 3 ' ' 7 0 . 0 1 2

2 0 . 0 0 2 8 . 0 . 0 1 8

3 0 . 0 1 1 9 0 . 0 1 0

4 0 . 0 2 4 10 0 . 0 2 0

5 ' 0 .0 1 5 11 ■ 0 . 0 2 0

6 0 . 0 0 4 12 . : 0 . 0 1 1

M ean and SD 0 . 0 1 2 ± 0 . 0 1 2 Mean and SD 0.015'± 0.007 between the results of the two groups and the overall mean is 0.013 ± 0.013pc/ 1.5 1 (an average daily urinary excretion. Pulse-height analyses of six electrodeposited sources (samples 7-12) were made individually, using a pulse-height analyser with a solid-state detector. The separate spectra were added to give a cumulative spectrum and the spectrometer background for the same period subtracted to give a difference spectrum. In Fig. 5, this spectrum is shown as a histogram, plotted in groups of 10 channels. The highest counting rates occur in the region between 4.2 and 5.4 MeV, which corresponds to the plutonium channel, but about 60% of the pulses fall outside this region, so that a useful reduction in the "blank" level is obtained. Regarding the nature of this "blank" activity, little can be deduced be­ cause the couting rates are so low. Several important naturally-occurring alpha em itters, notably Pc?10(5.3 M eV), Ra226(4.78 MeV) and Th228(5.4 MeV),

Fig. 5 ,

Alpha spectrum of "blank" samples obtained from unexposed personnel ALPHA-EMITTING PLUTONIUM IN URINE 273 fa ll in the range of energies covered by the plutonium channel. . In view of the fact that Pu239 has been detected in the lungs of non-.occupationally-exposed people [20, 21] from the retention of inhaled plutonium of fallout origin, it seems possible that such people can acquire a small systemic burden and excrete very low levels of plutonium in urine, which might account for some of this activity. .

4. CONCLUSIONS ' " '

The studies described in this report show that wet oxidation, followed by precipitation and chloroform extraction of plutonium-iron cupferrides, is the most satisfactory of the-various methods investigated for the concen­ tration of plutonium from urine and its subsequent purification. The plutoni­ um can then be electrodeposited quickly and quantitatively from an acid ammonium sulphate electrolyte and the overall recovery of 84 £ 7% is high and consistent. The alpha-emitting plutonium in the electrodeposited source can be determined by nuclear-track autoradiography, conventional scintil­ lation counting, or by the use of a solid- state counter developed for this purpose at AERE. The solid-state counter has a very low inherent back­ ground and the prototype counter, developed at AERE for plutonium bio­ assay measurements has three variable-level discriminators and may be used to count only those alpha particles within a predetermined range of energies. In this way, the plutonium can be counted with little loss in effi­ ciency, while effecting considerable reduction both in the inherent counter background and the "blank" activity. It is considered that even in its present state of development, this type of counter has advantages over autoradio­ graphic and scintillation counting methods. The cost of the equipment is unlikely to exceed that of the scintillation counter and the tedious manual counting of tracks required by the autoradiographic method is avoided.

ACKNOWLEDGEMENTS

The authors gratefully acknowledge the assistance of members of the Electronics Division at AERE in particular Mr. A. H. Muston, who designed and built the prototype solid-state counter described, and Mr. F. Allsworth who measured the alpha spectra of many samples. In addition we should like to thank Dr. J.C. Dalton of the Chemical Services Department at Windscale, who gave much help and information in the early stages of this work, and Mr. N. Taylor of the Health Physics Division, AWRE, Aldermaston, who has made several helpful suggestions.

REFERENCES .

[1] RECOMMENDATIONS OF THE INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION. Report of Committee II.on permissible dose for internal radiation, Pergamon Press (1959). [2] LANGHAM, W.H., Hlth Phys. 2 (1959) 172. . 274 F. J. SANDALLS and A. MORGAN

[3] SMALES, A. A ., AIREY, L ., WALTON, G. N. and BROOKS, R.O.R. AERE-C/R 533 (1950). [4] BROOKS, R .O .R ., AERE-AM 60 (1 9 6 0 ). [5] FARABEE, L.B., Clinton Laboratory Mon.-218 (1947).

[ 6] WEISS, H.V. and SHIPMAN, W .H., Anal. Chem. 33 (1961) 37. [7] DALTON, J.C., PG Report 284 (W) (1962). . - .

[ 8] BAINS, M .E.D., AEEW-R 292 (1963). [9] FARABEE, L.B ., ORNL-3492 (1963). [10] SCHWENDIMAN. L. C. and HEALY, J.W ., Proc. 2nd UN Int. Conf. PUAE 23 (1958) 144.

[11] SANDERS, S .M . and LEIDT, S. C . , Hlth Phys. 6 (1961) 189. [12] JACOBSEN, W.R., ANL-6637 (1961). ■” [13] KEMP, D.M ., AERE-R 4119 (1962). [14] SCHWENDIMAN, L.C ., HEALY, J.W. and REID, D. L., HW-22680 (1961). [15] MITCHELL, R. F., Anal. Chem. 32 (1960) 326. [16] LOVERIDGE, B. A., AERE-R 3266 (1960). . ' [17] DUPZYK, I. A. and BIGGS, M. W. TID-7616 (1960). ; [18] SANDALLS, F .J., and MORGAN, A., AERE-R 4391 (1964). [19] MUSTON, A. H. AERE-R 4539 (1964). ' [20] KREY, P. W ., BOGEN, D. and FRENCH, E .. Nature 195 (1962) 263. [21] OSBORNE, R .V ., Nature 199 (1963) 143. A DOUBLE-FILTER DEVICE TO MEASURE RADON AND THORON IN THE BREATH

W . JACO BI HAHN- MEITNER- INSTITUT FÜR KERNFORSCHUNG BERLIN, ABTEILUNG STRAHLENPHYSIK; BERLIN- WANNSEE, FEDERAL REPUBLIC OF GERMANY . /

Abstract — Résumé — Аннотация — Resumen

A DOUBLE-FILTER DEVICE TO MEASURE RADON AND THORON IN THE BREATH. The measurement of radon and thoron in the breath is a well-known sensitive method to estimate the radium- and thorium body burden in man. However, at low concentrations the conventional Rn- and Tn-counting methods are rather difficult in application. Compared with these methods, the described new method is distinguished by its simplicity and sensitivity as well as having the advantage of being able to determine low Tn-concentrations in the presence of Rn. ■ The principle of this method is as follows: The exhaled air streams through a fibrous .filter of high effi­ ciency for Rn- and Tn-decay products into a chamber, where new decay product atoms are formed by the decay of Rn- and Tn-atoms. These fresh build-up decay products are filtered out from the streaming air by a second filter at the air outlet of the chamber. The activity of the outlet filter is directly proportional to the Rn- and Tn- concentration in the exhaled air and can be easily measured by conventional filter- counting techniques. Devices using this principle are described, which allow the simultaneous measurement of Rn- or Tn-

concentrations of 100 pc/m 3 in the breath with an error below 10%. The Tn-sensitivity is independent of the Rn- con cen tration in the air.

DISPOSITIF A DOUBLE FILTRE POUR DOSER LE RADON ET LE THORON PRÉSENTS DANS L'HALEINE. Le dosage du radon et du thoron présents dans l'haleine constitue une méthode sensible bien connue pour déter­ miner la charge corporelle de radium et de thorium. Toutefois, pour de faibles concentrations, les méthodes classiques de comptage de Rn et de Tn sont d’application assez délicate. La nouvelle méthode décrite par les auteurs se distingue de ces méthodes par sa simplicité et sa sensibilité et elle présente également l'avantage de permettre de déterminer de faibles concentrations de Tn en présence de Rn. Cette méthode repose sur le principe suivant: l’air expiré passe, à travers un filtre de fibres retenant bien les produits de la désintégration de Rn et de Tn, dans une chambre où se forment de nouveaux produits de la désintégration des atomes de Rn et de Tn. Ces nouveaux produits de désintégration sont retenus par un second filtre, lorsque l'air sort de la chambre. L'activité du filtre de sortie est directement proportionnelle"à la concentration en Rn et en Tn de l'air expiré et peut être facilement mesurée au moyen de méthodes de comptage classiques. Les auteurs décrivent des dispositifs dont le fonctionnement est fondé sur ce principe et qui permettent de mesurer simultanément des concentrations en Rn et en Tn de 100 pc/m 3 dans l'haleine avec une marge d’erreur inférieure à 1 0 % La sensibilité de la mesure de Tn est indépendante de la concentration de l'air en Rn.

ПРИБОР С ДВОЙНЫМ ФИЛЬТРОМ ДЛЯ ИЗМЕРЕНИЯ СОДЕРЖАНИЯ РАДОНА И ТОРОНА В ВЫДЫХАЕМОМ ВОЗДУХЕ. Измерение радона и торона в выдыхаемом воздухе считается чувствительным методом оценки содержания радия и тория в организме человека. Однако при низких концентрациях стандартные методы счета радона и торона трудно исполь­ зовать По сравнению с этими методами предлагаемый новый метод отличается простотой и чувствительностью, а также возможностью определять малые концентрации торона в при­ сутствии радона. Сущность метода заключается в следующем: выдыхаемый воздух проходит через во­ локнистый фильтр, обладающий высокой эффективностью в отношении продуктов распада ра­ дона и торона, в камеру, где при распаде атомов этих элементов образуются новые продукты атомного распада. Эти вновь образовавшиеся продукты распада фильтруются из проходящего воздуха с помощью второго фильтра у выходного отверстия камеры. Активность выходного

275 276 W. JACOBI

фильтра прямо пропорциональна концентрации радона и торона в выдыхаемом воздухе и может быть легко измерена с помощью стандартных методов счета фильтра. Дается описание прибора, использующего этот принцип и позволяющего одновременно

измерять содержание радона и торона при концентрации 10 0 мкмккюри/м 3 в выдыхаемом воз­ духе с погрешностью ниже 10%. Чувствительность к торону не зависит от концентрации ра­ дона в воздухе.

DISPOSITIVO DE DOBLE FILTRO PARA MEDIR EL RADÓN Y EL TORÓN CONTENIDOS EN EL ALIENTO. La determinación cuantitativa del radón y del torón contenidos en el aliento constituye un método bien conocido y sensible para medir la carga corporal de radio y de torio en el'hombre. Sin embargo, cuando las con­ centraciones respectivas son bajas los métodos corrientes de recuento del Rn y del Tn son de aplicación bastante difícil. En comparación con ellos, el nuevo procedimiento que se describe en la presente memoria se distingue por su sencillez y sensibilidad, asf como por la ventaja de permitir determinar bajas concentraciones de Tn en presencia de Rn. El fundamento de este procedimiento es el siguiente: el aire exhalado atraviesa un filtro fibroso que retiene muy eficazmente los productos de desintegración del Rn y del Tn, y pasa a una cám ara en la que se forman nuevos átomos' de productos de desintegración por transformación de los átomos de Rn y de Tn. Estos nuevos productos de desintegración son separados del aire exhalado por un segundo filtro situado a la salida de la cámara. La actividad del filtro de salida es directamente proporcional a la concentración del Rn y del Tn en el aire exhalado y puede determinarse fácilmente por cualquier procedimiento corriente de recuento de la radiactividad de filtros. El autor describe dispositivos basados'en éste principio, que permiten medir simultáneamente concen­ traciones del orden de 100 pc/m 3, del Rn o del Tn contenidos en el aliento, con un error inferior al 10%. La sensibilidad al Tn no depende de la concentración del Rn en el aire. ' •

INTRODUCTION ' .

The measurement of Rn222 (Rn) and Rn220 (Tn) in the breath is a well- known method for estimating the burden of Ra226 and Th232 or Ra228 in the human body. However, the methods used hitherto for the measurement of low Rn and Tn concentrations in the breath, whichuse compensated or pulsed ionization chambers [1],. a-cavity-scintillators [2] or adsorption on activated charcoal [3', 4, 5, 6], are rather difficult in application. On the basis of a "double-filter"-principle BLANC et al. and the author have recently developed new methods for the measurement of Rn and Tn in flowing’gases [7, 10, 11] . In a simple way this double-filter method allows a sensitive, continuous measurement of Rn and Tn in air; furthermore, there is an advantage in that small Tn concentrations can be determined in the presence of Rn. The double-filter method is therefore well suited for the study of the Rn and Tn distribution in the atmosphere and was already applied in this field [7, 10, 11]. On account to its simplicity and high sensitivity, this method w ill also be useful for the measurement of Rn and Tn in the breath. In this paper the feasibility of this double-filter method for this purpose is discussed. In the first section the principle of the method is explained; in the following sections special devices of this type are described which are suited for the determination of Rn and Tn in the breath.

1. PRINCIPLE OF THE DOUBLE-FILTER METHOD ■'

The principle of the double-filter method is shown schematically . in Fig. 1. The Rn- or Tn-containing air flows through a-fibrous filtefr (inlet DOUBLE-FILTER DEVICE FOR RADON AND THORON 277

Filter - arrangement . (o = Rn.Tn-atom, * = Decay-product-atom)

' .'I Filter . 2. Filter

Filter activity. 1. Filter Aí ~ aop (0) • v .

2. Filter 3 dp ( L) • v ~ а Вп< în (0) • f (ф, L,v) ' Fig. 1 '

Principle of the double-filter method with a fibrous second filter filter) into a chamber and is sucked off at the other side of the chamber through a second filter (outlet filter). The inlet filter has a high efficiency and retains the aerosol and the Rn and Tn decay products p rim arily existing in the air, whereas the noble gases Rn and Tn pass this filter undisturbed. In the decay chamber between both filters Po218 (RaA) atoms or Po216 (ThA) and Pb212 (ThB) atoms are formed by the decay of Rn or Tn atoms. A fraction of these new decay products diffuse to the chamber walls, where they are deposited. The rest remains in the air and is filtered out by the second filter at the air outlet of the decay chamber. The enrichment of these fresh decay products on the outlet filter follows the same law as in the case of the conventional one-filter method; as their concentration in the air of the decay chamber is proportional to the Rn or Tn concentration, the activity of the outlet filter is directly proportional to the Rn or Tn .concentration of the air at the inlet of the device. Therefore the Rn or Tn concentration of air can be easily determined by-the measurement of the activity on the outlet filter during or after sampling. If both emanations are present, the fraction of Rn and Tn can be easily determined by analysis of the decay curve of the outlet-filter activity after the end of sampling. ■ By this double-filter method the difficult measurement and determination of low Rn and Tn levels in air is converted to a simple filter measurement, which can be done with a low background and conventional a-counting'tech- niques. It must be emphasized, that the Rn and Tn determination is not influenced by their decay products in the air sucked into the double-filter device, because these decay products are retained in the inlet filter; the measurement of the inlet-filter activity gives therefore a simultaneous indi­ cation "of the RaB (or RaC) and ThB (or- ThC) content of the air. 278 W. JACOBI

The sensitivity of this double-filter method depends on the number of Rn or Tn atoms decaying in the decay chamber and the fraction of decay products which are deposited on the chamber walls. In the lower part of Fig. 1 the variation of the concentration of Rn, Tn and their decay products in the air-flow through the device is shown schematically. Behind the inlet filter the concentration of decay products increases from zero and reaches a saturation value, when the production rate of these atoms is compensated by their loss-rate on account of diffusion to the chamber walls. To test these expectations in preliminary studies narrow tubes were used as decay chambers and the RaA/Rn ratio in the air at different distances behind the inlet filter was measured. Figure 2 shows the results for different

F i g .2

Build-up ofRaAin air streaming through a pipe after préfiltration at L=0

tube diameters, ф, and flow rates, v. As expected the RaA/Rn ratio in­ creases behind the inlet filter according to an exponential law of the form

: (“ir-)L = (iir i x [1-exp (-const-L)i

(a = specific activity, L = tube length). In tubes of this small diameter the loss rate by diffusion to the walls is high and the observed ratio (aRaA /aRnL is therefore far below radioactive equilibrium. Figure 2 shows that this ratio increases with increasing tube diameter and decreasing flow rate. As the activity of the outlet-filter is proportional to aRaA X v, the maximum activity on the outlet filter occurs at a definite flow rate. It follows from these experiments, that the volume of the decay chamber should be large to reach a high RaA/Rn ratio in the air at the outlet filter. However in the, case of Tn its rapid decay with a half-life-time of 54. 5 s sets an upper limit for the ratio V/v (V = Volume of the decay chamber). DOUBLE-FILTER DEVICE FOR RADON AND THORON 279

2. A DOUBLE-FILTER DEVICE WITH FIBROUS OUTLET FILTER

Several double-filter devices for the determination of low Rn and Tn concentrations in air were developed. Figure 3 shows a device in which the o-activity of the outlet filter is continuously measured during sampling.

PRINTER SCALER

F i g .3

Double-filter device with fibrous outlet filter for the continuous measurement of Rn and Tn in flowing air

The decay chamber consists of a cylindrical tube, 25-cm diam. , made of polyvinylchloride. The length of the cylindrical part of the decay chamber can be varied to 50 cm, 100 cm and 150 cm. The corresponding.total volume of the decay chamber is 26 1, 51 1 and 75 1. The perforated front wall carries the inlet filter of 25-cm diam., which has a filter efficiency for Rn and Tn decay products of > 99. 95%. The air is sucked off on tlie other side of the chamber through an outlet filter of 5 cm effective diameter, which is easily exchangeable and is of the same type as the inlet filter. At a variable distance above the surface of the outlet filter a ZnS-scintillator of 3-in diam. is mounted. Normally the width of the- air gap between the filter and scintil­ lator is about 2 mm, which gives a detection efficiency of 45% for the or-particles of RaC1 and ThC + ThC1. The output pulses of the attached photo­ multiplier are fed to a counting device and are recorded either with a line- recorder or a printer. The Rn and Tn sensitivity of the device was determined by sweeping air through calibrated RaCl2 and Th (N03)4 solutions, the latter one containing 30 yr-old thorium. The emanated Rn and Tn were continuously inserted into the air flow before the inlet filter of the device, having regard to the Tn decay in the dead volume between the emanating solution and the inlet filter. The resulting specific activity in the air entering the decay chamber was varied for Rn from 0. 5 - 500 pc/1 and for Tn from 10 - 1000 pc/1. In the following the ratio between the a-activity on the outlet filter at equilibrium (on the filter) to the specific activity of the air at the inlet of the decay chamber is called differential sensitivity, s, of the device. The integral S = s dt is called the integral sensitivity and gives the total number of counts from the beginning of sampling (t = 0) till the end of the sampling period t. - At constant flow rate the differential and integral sensitivity were inde­ pendent of the specific activity. The dependence upon flow rate for Tn is 2 80 W . JACOBI

‘ ‘ Fig.4 ‘ ' ■

Differential Tn sensitivity of double- filter devices with fibrous outlet filter plotted as a function of , the air-flow rate v (Parameter: Decay chamber volume V) ' ' shown in'Fig. 4 for the three decay chambers of diffèrent length but equal diameter. On the upper scale of the abscissa the corresponding.mean resi­ dence time V/v of the1 air inthe chamber is recorded. ... The- Tn sensitivity of these double-filter devices reaches a maximum at V/v = 0. 5 min. With decreasing flow r.ate the Tn sen sitivity decreases strongly, because then most of the Tn atoms decay in the first part of the chamber and a large fraction of the form ed ThB atoms- diffuse to the walls. , For most purposes the 51-1 chamber was used. At a ,flow rate of v: = ;2. 5 m3/h the differential Tn sensitivity of this device is . . ...

. s.j... = 7.0 ± 0. 5 cpm -l/pc ■ : .

(détector efficiency 45%)'. The corresponding differential Rn sensitivity was determined to be . : . sRn =.19 ± 1 cpm-l/pc. . .

The mean background counting rate is 0. 2 cpm. At equilibrium on the outlet filter therefore O.'l pc/1 of Rn or Tn in air, which is approximately the mean specific activity of these nuclides in atmospheric air near grdund level, can be measured in a counting period of 1 h with a mean statistical error' of 10% in the case of Rn and of 20% in the case of Tn. DOUBLE-FILTER DEVICE FOR RÀDON AND THORON 281

A double-filter device of this type is therefore well'suited for the con­ tinuous measurement of Rn in atmospheric air and enables a simple quantita­ tive determination of the poorly known Tn content of atmospheric air [7] to be made. As is shown later, a device of this type is also suitable for the measurement of very low Rn concentrations in breath. The sensitivity of this device is limited for thé detection of Tn in breath because the breathing rate is only 5-10 I/min. At v = 5 1/min the differential Tn sensitivity is sTn =0.5 cpm-l/pc (see Fig. 4). ■ ■

3. A DOUBLE-FILTER DEVICE WITH ELECTROSTATIC FILTRATION

Another double-filter device which has been developed for the measure­ ment of Rn and Tn in breath, makes use of the well- known fact that a fraction of the RaA or ThA and ThB atoms - originating by the a-decay of Rn or Tn atoms - are positive ions at the beginning óf their lifetime. In this device the fibrous filter at the outlet of the decay chamber is replaced by an electro­ static filter in the decay chamber. The positive charged decay-product atoms originating in the decay chamber are collected by the electric field on the cathode, whose a-activity is continuously measured. As the number of collected ions is proportional to the decay rate of Rn or Tn in the decay chamber, the activity on the collector electrode should be directly pro­ portional to the specific Rn or Tn activity of the air flowing through the chamber...... ! A device of this type is shown in Fig. 5. After passing the inlet filter, which retains the aerosol and the decay products of Rn and Tn initially in the air, the air flows into a cylindrical decay chamber of 10-1 volume, made of polyvinylchloride. The positive high-voltage electrode is placed at the front wall and the negative, grounded collector electrode is placed in the centre of the opposite wall. The collector electrode is an Al-covered, cir­ cular polyester foil of 40 mm diam. with a thickness of only 0. 9 - 1 mg/cm2 to prevent the absorption of a-rays in the foil. It can be easily replaced

SCALER PRINTER

F ig .

Double-filter device with electrostatic filtration for the continuous measurement of Rn and Tn in air 282 W. JACOBI

if necessary. A ZnS scintillator of 50 mm diam. is attached to the back of the collector foil and measures the a-activity of the collector during sam pling. The sensitivity of this device was checked in the same way as for the device with a fibrous outlet filter. As expected it is independent of the air specific activity but varies with the precipitation voltage up as is plotted in Fig. 6. The maximum sensitivity is reached at up = 500 Volt. At up = 500 V the differential Rn sensitivity is

" sRn = 6 ± 0. 2 cpm-l/pc .

independent of the air-flow rate. - . The Tn sensitivity changes with the flow rate on account of the short half-life of Tn. Figure 7 shows the increase in the net counting rate and the number of total counts after the' injection of 1500 pc Tn/min at a flow rate of 5 1/min and a precipitation voltage of up = 500 Volt. From Fig. 7 can be calculated a differential Tn sensitivity of

1 sTn= !• 2 ± 0. 1 cpm-l/pc

and an integral Tn sensitivity of "

STn (20 min) = 0. 80 cpm-l/pc

STn(lh) = 3.8 ■ "

' STn(3h) =18 "

STn (10 h) = 170 "

The background of the detector is 0. 2 cpm. ’

The Tn sensitivity is more than a factor 2 higher than that of the 51-1 device with a fibrous outlet filte r at a flow rate of 5 1/min. The device with electrostatic filtration is therefore better adapted for the measurement of Tn in the breath. . •

4. FEASIBILITY OF DOUBLE- FILTER DEVICES FOR THE MEASUREMENT OF RADON AND THORON IN THE BREATH 4

Rn and Tn content of the breath

The mean elimination rate E and specific activity a of Rn and Tn in breath, on account of deposits of their parent nuclides in the human body, is given by

E (pc/min) = 6. 0 X 101 X X (sec"1 ) X e X q (цс)

a (pc/1) = E (pc/min)/vb( 1/min) . . DOUBLE-FILTER DEVICE FOR RADON AND THORON 283

PRECIPITATION VOLTAGE (kV) '

F ig . 6

Differential Rn sensitivity of the double- filter device with electrostatic filtration in a 10-1 chamber plotted as function of the precipitation voltage

TIME (h )

Fig-1 '

Response of the double-filter device with electrostatic filtration in a 10-1 chamber versus time at injection of 1500 pc Tn/min with a flow rate of v = 5 1/min where

q = the body burden of the parent nuclide, ' . e = the emanation efficiency of the body, X = the radioactive decay constant (of Rn or Tn), vb ~ the breathing rate. 284 W . JACOBI

The emanation efficiency is about 0. 7 for Rn [8] and about 0. 1 for Tn [5]. The resulting mean elimination rate and specific activity of Rn and Tn in the breath for the maximum body burden of Ra226 and Th232 or Ra228 are given in Table I.

TA BLE I

MEAN ELIMINATION RATE AND SPECIFIC ACTIVITY OFRn AND Tn IN BREATH AT THE MAXIMUM PERMISSIBLE BODY BURDEN OF R a226 AND Th232 OR R a228

M axim u m E xp ired Elimination rate Spec, activity* permissible n u clid e (p c /m i n ) ( p c / l ) body burden

0 . 1 tic R a226 'R n - 8 . 8 1 . 8

0 . 0 4 p c T h 232 Tn' ' 3 0 0 0 6 0 0

* Assuming a breathing rate.of vb = 5.1/min.

The mean specific activity in ordinary room air is about 0. lpc/1 radon and about 0. 2 pc/1 thoron. To m easure a sm all Ra226 body burden, the subject must therefore inhale aged tank air with a low Rn and Tn content for a sufficient period of time before the breath sample is taken.

Rn measurement in thebreath '

The arrangement of a double-filter device for the measurement of Rn in breath is shown in Fig. 8. The subject exhales through the outlet valve, of a face mask into’a spirometer and the breath rate. is. measured. Then a fraction of this sample is transferred to the evacuated decay chamber of the device, until atmospheric pressure is reached. In the device with a fibrous outlet filter (described in section 2) the air is now pumped at a definite flow- rate in a continuous circuit through the decay chamber. In the case of the electrostatic filter device (described in section 3) an air flow through the chamber is not necessary because its Rn sensitivity is independent of the flow rate. The equilibrium activity on the outlet filter is measured as long as necessary to get a small statistical counting error. If RRn is the measured mean value of the net counting rate due to radon at equilibrium on the outlet filter, the Ra226 body burden can be calculated from the relation

. vb (.1 /min) X RRn (cpm) ^Ra ^ c > = ------88 x s«-fe) "

where SRn is the differential Rn sensitivify of the device at the adjusted flow rate or precipitation voltage in the decay chamber. In Table II the resulting counting rates at different Ra226 body burdens are presented-for the double DOUBLE-FILTER DEVICE FOR RADON AND THORON 285

SPIROMETER ’ ' only necessary for the device

Arrangement of a double-filter device for the measurement of Rn in the breath filter devices described. In the last column is listed the statistical standard deviation for the determination of the body burden. ■ From Table II it follows that 10% of the maximum permissible body burden of Ra226 can be measured during a counting period of 1 h with a standard deviation of about 10%, and 1% of the maximum permissible body burden can be measured during 3 h with an error of about 20%.

Tn-measurement in the breath -

On account of its short half-life of 54. 5 s, Tn must be measured directly in the expired air flow. Figure 9 shows an arrangement which can be used for this purpose! The air is expired into a valved face mask. A known fraction of the expired air passes into a small spirometer and is sucked off by a pump through the decay chamber of the double-filter device. The spiro­ meter converts the pulsed flow into a constant and adjustable flow to obtain a defined sensitivity of the double-filter device. Normally the air flow through the decay chamber is adjusted to v = 5 1/min. ' The excess expired air is blown off through a regulator valve and a gas meter to define the total breathing rate. ' ■ .. For a short breathing period the total number of counts from beginning to end of the breathing period is measured and recorded. If N is the total number of net counts measured during the breathing period t and Бтп (t) is the corresponding integral Tn-sensitivity of the double-filter device, the body burden of Th232 or its daughter nuclides Ra228 and Th228 results in

' v b (1/m in) X N (counts) q Th0,c) = 1.3 X lo"5 X exp (XTnX Vd /v) X (counts. 1/p¿)-

The factor exp (ATnVd/v) allows for the decay of Tn in the "dead-volume" Vd between the face mask and thé inlet filter of the double-filter device. V^ is mainly the measured mean volume of the spirometer during the breathing period. Normally this correction factor lies in the range 0. 5 - 0. 7. In Table III the resulting Tn response of the electrostatic filter device is pre- 8 W JC BI JACO W. 286

• TABLE II

RESPONSE OF DOUBLE-FILTER DEVICES TO RADON IN THE BREATH DUE TO 100%, 10% AND 1% OF THE MAXIMUM PERMISSIBLE Ra226 BODY BURDEN

T y p e Differential • C o u n tin g Ra226 Mean detector response S tan d ard o f Rn sen sitiv ity period body burden (c o u n ts ) d e v ia tio n

d e v ic e ( c p m - 1/ p c ) ( h ) (№ ) R adon B ack ground (°/o)

0 . 1 2 0 4 0 ■ 12 2 . 2

1 0 . 0 1 2 0 4 1 2 7 . 1 . 51-1 chamber 19 0.001 2 0 . 4 12 3 3 with fibrous - (at v = 2. 5 m 3/ h ) 0 . 1 6 1 2 0 3 6 . 1 . 3 outlet filter 3 0 . 0 1 6 1 2 . 3 6 ' 4 . 2

0.001 6 1 . 2 3 6 1 6

0 . 1 8 4 0 12 3 . 5

1 0 . 0 1 ' 8 4 12 13

1 0 - 1 c h a m b e r 7 . 6 0.001 8 . 4 12 68

(at up = 500 V) 0 . 1 2 5 2 0 3 6 2 . 0 static filtration

3 0 . 0 1 2 5 2 3 6 6 . 7 ■

0.001 2 5 . 2 3 6 31 DOUBLE-FILTER DEVICE FOR RADON AND THORON 2 87

EXPIRED AIR V„p = Vinh

' F ig . 9

Arrangement.of a double-filter device for the measurement of Tn in the breath

TA BLE III

RESPONSE OF THE ELECTROSTATIC FILTRATION DEVICE TO THORON IN THE BREATH DUE TO 100%, 10% AND 1% OF THE MAXIMUM PERMISSIBLE BODY BURDEN OF Th232 (decay-chamber volume 10 1; flow-rate 5 1/min; p recip ita tio n v o lta g e 500 V; differential Tn sensitivity 1. 2 cpm-l/pc)

B reath in g In te g ra l T h 232 M ean det ector response Standard period Tn sensitivity body burden (co u n ts) ‘ d e v ia tio n

( c o u n t s - 1/ p c ) (ДС) T h oron B ack grou n d

0 . 0 4 4 9 0 4 . 4 . 5

2 0 m in 0 . 8 0 . 0 . 0 0 4 4 9 4 1 5

0 . 0 0 0 4 4 . 9 4 6 0

0 . 0 4 2 3 4 0 1 2 ■ 2 . 1

1 h 3 0 . 0 0 4 2 3 4 1 2 6 . 7

0 . 0 0 0 4 2 3 . 4 .12 2 5

0 . 0 4 1 1 0 0 0 3 6 1 . 0

3 h . 18 0 . 004' 1 1 0 0 3 6 3 . 1

• 0 .0 0 0 4 1 1 0 36 11 ' sented at different Th232 body burdens for breathing periods of 20 min, 1 h and 3 h (calculated with Vd = 0). 288 W . JACOBI

From Table III it follows that 10% of the maximum permissible Th232 body burden can be measured with a mean statistical error of 15% during a breathing period of 20 min and 1% of the maximum permissible body burden can be measured with a deviation of 25% in 3 h. The influence of Rn can be recognized easily from the decay curve at the end of breathing, but under normal conditions it is negligible. ' ; ,

5. CONCLUSIONS . .

The applications of the double-filter methods described indicate that this method allows the simple detection and discrimination of Rn and Tn down to concentrations of about 10 pc/m3. The method gives high sensi­ tivity with simplicity and stability. The technical effort is low and only con­ ventional a-counting techniques are required; traps for the adsorption of water vapour and carbon dioxide are not necessary. The.method is well suited for the measurement of Rn and Tn in the breath and enables the detection of 1% of the maximum permissible body burden of Ra226 and Th232 or its daughters Ra228 (Ms Th 1) and Th22S (RdTh) in a convenient .time period. The sensitivity of this method for the detection of these nuclides in the body is comparable with that of good whole-body monitors [9]. Also, for the first time, continuous measurement of Rn in atmospheric air and a simple quantitative determination of the poorly known Tn-content of atmospheric air can be made by this method. ! . .

REFERENCES .

[1] RUNDO, J., WARD, A.H. and JENSEN, P.G ., Measurement of thoron in the Ureath, Phys. Med. Biol. 3 (1958) 101-110. [2] LUCAS, H.F.., Improved low-level alpha-scintillation counter for radon, Rev. Sci. Instr. 28 (1957) ■ '680-683. - ■. . _ Í [3] HURSH, J.B ., Measurement of breath radon by charcoal adsorption, Nucleonics, No.l,_12 (,1954) 62-65. [4] STEHNEY, A .F., NORRIS, W .P., LUCAS, H.F. and JOHNSTON, W .H., A method for measuring the rate of elimination of radon in breath, Amer. J. Roentgenol., Rad. Therapy anil Nuclear Med. 73 (1955) 7 8 5 -8 0 2 . _ [5] HURSH, J.B . and LOVAAS, A ., A device for measurement of thoron in the breath, Health Physics 9 (1 9 6 3 ) 6 2 1 - 6 2 7 . ' .

[ 6 ] HURSH, J.B ., Feasibility study of a new device to measure radon in the breath, Symposium on Radio­ logical Health and Safety in Mining and Milling of Nuclear Materials, IAEA, Vienna,-Vol.II (1964) p . 4 5 1 . ■ [-7] JACOBI, W ., A new method to measure radon and thoron in flowing gases and its use to determine the thoron content of atmospheric air, Int. Symposium on Radioactive Pollution in Gases, Saclay (Nov. 1963) in: ' Comptes Rendu, in press. , _

.[ 8 ] NORRIS, W .P .., SPECKMAN, T .W . and GUSTAFSON, P . F ., Studies on the metabolism of radium in man, . -Amer. J. Roentgenol., Rad, Therapy and Nuclear Med. 73 (1955) 7S5-S02. ' [9] MEHL, H.G. and RUNDO, J ., Preliminary results of a world survey of whole-body monitors, Health Physics 9 (1963) 607-614. [10] BLANC, D ., FONTAN, J. and YEDRENNE, G. , «U n procédé de dosage continu du Radon dans l'air atmosphérique. Application à la prospection de l'uranium» J. Phys. Radium 21 (1960) 176-180. [11] FONTAN, J ., Le dosage des radioéléments gazeux donnant des produits radioactifs de filiation. Son application à la mesure de la radioactivité naturelle de l'atmosphère. (Thèse),Univ. de Toulouse, (fév. 1964). DOUBLE-FILTER DEVICE FOR RADON AND THORON 289

DISCUSSION

H. WYKER: I understand Dr. Jacobi took an emanation efficiency of 10% for thoron as a basis for calculating the Th232 body burden. As the half­ life of thoron is only 54 s - very short in comparison with the half-life of ràdon(3.8d) - this emanation efficiency seems very high as compared with that of radon. Could the speaker comment on this? W. JACOBI: The figure of 10% was taken from the literature, being derived from measurements on thorotrast patients. In these cases the thoron is generated mainly in the liver and spleen, and evidently can escape quickly to the blood and hence lungs. Radon, on the other hand, is generated by radium deposited mainly in the bone, and clearly diffuses to the blood much less quickly. J. RUNDO: What assumptions has Dr. Jacobi made concerning the state o f equilibrium between the Th232 and its daughters Ra228 arid Th228? I f the 10% emanation efficiency refers to thorotrast cases, then the measurement of thoron in the breath w ill give the'Th228 content of the liver and spleen, rather than the Th232 content of the whole body. W. JACOBI: This is quite true. In Table III radioactive equilibrium between Th232, Ra228 and Th228 in respect of the total body content of these nuclides was assumed. . N.I. SAX: Can Dr. Jacobi say something about the concentration of Th232 in the spleen as opposed to the liver? W. JACOBI: No. We have not measured the distribution in thorotrast patients. This paper deals only with the feasibility of the double-filter method for the measurement of radon and thoron in the breath. C.W. SILL: Have you compared the sensitivity of your gasometric measurements with that of urinalysis or faecal analysis?^ W. JACOBI: Not ÿet. But I believe that in the case of Ra226 the exhalation method allows a more reliable estimation of the body burden. H. WYKER: Does not the exhaled air which you measure contain thoron which has just been inhaled from the surrounding atmosphere? W. JACOBI: The mean thoron content of normal room air is about 0.2 pc/1. This is negligible compared with the thoron concentration in the breath of the cases we are measuring.

TOTAL COUNTING AND SPECTROSCOPY IN THE ASSESSMENT OF ALPHA RADIOACTIVITY . IN HUMAN TISSUES

W .V. MAYNEORD AND C.R. HILL PHYSICS DEPARTMENT, INSTITUTE OF CANCER RESEARCH, ROYAL CANCER HOSPITAL, CLIFTON AVENUE, BELMONT, SURREY, ENGLAND '

Abstract — Résumé — Аннотация — Resumen

TOTAL COUNTING AND SPECTROSCOPY IN THE ASSESSMENT OF ALPHA RADIOACTIVITY IN HUMAN TISSU ES. A seven-year programme of measurements of alpha radioactivity is briefly reviewed. The pro­ gramme which was based initially on a sensitive and simple counting technique, using thin zinc cadmium sulphide screens, now employs in addition techniques of alpha spectroscopy developed for work with very large area low-specific-activity sources. These techniques are capable of measuring specific activities down to 10 " 13 с/g and can in certain cases provide energy resolution of the order of 40 keV together with independent identification based on half-life measurements. Two alpha spectrometers are described which have been used on a wide variety of human and environmental materials. Particular attention has been given to the study of Po210, now known to be present in a variety of foods and human tissues. Techniques are also described for the study o f Pu239 in the atmosphere and examples given of .the results of measurements of low-level air contamination with this material.' Techniques for the measurement of Pu 239 in normal human tissues are described and some results given. Consideration is given to the contribution made by Po 210 to the natural background radiation dose-rate in human bone and reproductive organs. It is shown that this contribution may be a significant fraction of the total dose received.

DÉTERMINATION DE LA RADIOACTIVITÉ ALPHA DES TISSUS HUMAINS PAR DOSAGE DE.L'ACTIVITÉ GLOBALE ET SPECTROSCOPIE. L'auteur fait brièvement le point de l'exécution d’un programme septennal de mesures de la radioactivité alpha. Au début, on procédait seulement à un dosage simple et sensible au moyen d’écrans minces de sulfure de cadmium et de zinc; maintenant, ce procédé est complété par l'emploi de méthodes de spectroscopie alpha qui ont été mises au point à l'occasion de travaux sur des sources à faible activité spécifique et à très grande surface. Cés méthodes permettent de* mesurer des activités spécifiques aussi faibles que 10 -13 c/g , d’obtenir, dans certains cas, une résolution en énergie de l'ordre de 40 keV et, indépendamment, d’identifier les radionucléides d*après des mesures de la période. L'auteur décrit deux spectromètres alpha qui ont été utilisés avec des substances très diverses (tissus humains et échantillons du milieu ambiant). Il s'est attaché particulièrement au dosage du 210Po que Гоп sait maintenant être présent dans certains produits alimentaires et dans des tissus humains. Il expose les méthodes employées pour l'étude du 239Pu dans l’atmosphère et donne des exemples de mesures de faibles concentrations de ce radionucléide dans l’air. Il expose également~les méthodes utilisées pour le dosage du 239Pu dans les tissus humains normaux et communique quelques résultats. Il étudie la contribution du 210Po dans le débit de dose dû au rayonnement naturel dans les os et les organes de reproduction humains; Il montre que cette contribution peut constituer une fraction significative de la dose totale reçue.

ОБЩЕЕ ИЗМЕРЕНИЕ И СПЕКТРОСКОПИЯ ПРИ ОПРЕДЕЛЕНИИ АЛЬФА-РАДИОАКТИВ­ НОСТИ ТКАНЕЙ ЧЕЛОВЕКА. Дается краткий обзор программы по измерению альфа- радиоактивности. Выполнение программы, которое первоначально основывалось на использо­ вании простых и чувствительных методов подсчета с помощью тонких цинк-кадмий-сульфидных экранов теперь включает использование метода альфа-спектроскопии, разработанного для большого количества источников с низкой специфической активностью. С помощью этих ме­

тодов можно измерять уровни специфической активности вплоть до 1 0 -13 кюри/г и в некоторых случаях получать разрешение энергии порядка 40 кэв наряду с независимой идентификацией, основанной на измерении периода полураспада. Дается описание двух альфа-спектрометров, использованных при измерении большого количества образцов, взятых у людей, и материалов

окружающей среды. Особое внимание уделялось изучению Ро 210 , который, как известно, при­ сутствует в ряде пищевых продуктов и тканях человеческого организма. Описаны методы

291 292 W.V. MAYNEORD and C.R. HILL

исследования Pu 239 в атмосфере и приведены примеры результатов измерения загрязнения

воздуха малыми количествами этого материала. Описаны методы измерения Pu 239 в н ор­

мальных тканях у человека и приведены некоторые результаты. Обсуждается роль Ро 210 в повышении уровня естественного фонового облучения костей и репродуктивных органов че­ ловека. Показано, что эта доля может составлять значительную часть обшей полученной д о з ы .

EVALUACION DE LA ACTIVIDAD ALFA EN LOS TEJIDOS HUMANOS POR RECUENTO GLOBAL Y ESPECTROSCOPIA. Los autores describen sucintamente un programa de mediciones de la actividad alfa, que abarcó siete años. El programa, que se basaba inicialmente en el empleo de una sensible y sencilla técnica de recuento, con ayuda de pantallas delgadas de sulfuro de cinc y de cadmio, comprende en la actualidad técnicas de espectroscopia alfa especiales para fuentes de gran superficie y de baja actividád es­ pecífica. Estas técnicas permiten medir actividades específicas de hasta 10* 13c/g y pueden en ciertos casos alcanzar una resolución energética del orden de 40 keV, permitiendo al mismo tiempo una identificación independiente basada en la medición del período de semidesintegración. Se describen dos espectrómetros alfa utilizados para examinar una gran variedad de materias provenientes del cuerpo humano y del medio ambienté. Se ha dedicado especial atención al estudio del 210Po que, como ahora se sabe, se halla presente en buen número de productos alimenticios y de tejidos humanos. Asimismo, se describen técnicas de análisis del 239 Pu presente en la atmósfera y se citan ejemplos de los resultados de las mediciones de aire ligeramente contaminado por esa sustancia. Se describen asimismo procedimientos de medición del 239 Pu en tejidos humanos normales y se exponen algunos resultados. Se estudia la contribución aportada por el 210Po a la dosis de irradiación natural proveniente del medio ambiente en el esqueleto y en los órganos reproductores del hombre. Se demuestra que esa aportación puede constituir una fracción significativa de la dosis total re cib id a .

1. INTRODUCTION

There is increasing awareness of the possible significance of the dose received by human tissues from alpha-ray-emitting materials. Whereas a few years ago the emphasis was almost solely upon radium and its daughter products known to be present in bone, the issue has now become a wider one and there is a real possibility that alpha emitters, even in soft tissues, contribute an appreciable component. ; Our knowledge of alpha-ray levels is relatively recent and it is only a few years since estimates of alpha-ray emitters as obtained by different investigators differed by two orders of magnitude ; There are two possible approaches to evaluation of these levels. We may measure either the total alpha activity, or that of a specific and par­ ticular nuclide. The former approach seems to occur naturally to physicists and the latter to chemists. After some years of experience we have come to the conclusion that both approaches are necessary.

2. MEASUREMENT OF TOTAL ALPHA ACTIVITY

If we turn to the measurement of the total activities,- it seems that there is still no serious rival to zinc sulphide as the detecting clement for alpha particles in spite of much work on the possible use of plastic scintillators. Some investigators have preferred the use of a proportional counter thus avoiding the electronic complexities associated with the use of scintillation counting devices incorporating photomultipliers. ALPHA RADIOACTIVITY IN HUMAN TISSUES

F ig . 1

Construction of capsule used for total alpha counting

One of the main problems in the use of zinc sulphide, or indeed any other scintillating material, is its eventual contamination to a degree which becomes of importance in the measurement of the low activities en­ countered in the environment. We have attempted to overcome this diffi­ culty by sealing the material in intimate contact with its own zinc sulphide screen, F ig.l [1]. Such sealed capsules can be made up simply and cheaply and may be conveniently stored for measurements over long time intervals. We have found that these repeated measurements are of considerable im ­ portance in the identification of the materials present from their-half-lives. This arrangement makes possible identification of radium-226, for example, by virtue of the growth of radon-222 in the sealed capsule. ■ In the same way the presence of polonium-210 can often be detected in cases where radio­ active equilibrium has been disturbed. We might perhaps remark that radio­ active equilibrium is, in fact, rarely met with in the biosphere except for very short-lived materials following a long-lived parent. One of the interesting features of this technique is the opportunity it affords for the detection of "fast pairs" of alpha particles originating from radon-220 and polonium-216, or radon-219 and polonium-215. These "fast pairs" provide simple means for detecting the presence in a sample of the appropriate members of the thorium and actinium series. The total alpha technique is simple, quick and reliable, provided that care is taken in the design and construction of equipment so that high elec­ tronic stability is achieved. This, is of particular importance as the counting times are often long, in our own experiments 16 h is a common period of observation. With this technique it is possible to make measurements of specific alpha activities down to 10"13 с/g, which in practice covers almost all environmental materials. Simultaneously we may obtain information as to what groups of nuclides are present. We have used this technique to ob­ tain information on total alpha activity of human and animal tissues and in the human environment generally [2, 3]. The technique has also proved very valuable in studies of foodstuffs [4], drinking waters [5] and human excreta. The method is versatile in that it may be used for liquids and solids and, with some modification, for gases. . ' • Having measured the total alpha activity of a given material, the next and obvious step is to attempt a closer identification of the nuclides present. We may again proceed in one of two ways, obtaining either an overall picture of the nuclides present or, by chemical separation, making a careful quanti­ tative study of a particular nuclide. Both approaches have their advantages, 294 W.V. MAYNEORD and C.R. HILL the chemical method requiring very careful control if contaminating ma­ terials are not to be introduced from vessels or reagents. i

3. SPECTRO SC O PIC TECHNIQUES

If no chemical concentration is attempted., one is driven inevitably to large area alpha spectroscopy. With specific activities as low as 10"13 c/g the achievement of significant counting-rate and the' simultaneous avoidance of appreciable self absorption for alpha particles, requires the use of source areas of the order of 10 000 cm2, even in a device having high counting geo­ metry. It seems that only a pulse ionization chamber could accommodate sources of such a size. We have, therefore, designed and built an instrument of this type in which the source material (finely ground and slurried) is sprayed onto a 90X 180-cm sheet of 0.125-mm-thick aluminized "M ylar" or cellulose acetate, which can subsequently be rolled to form the outer electrode of a cylindrical ionization chamber (Fig. 2). .

Large area concentric cylindrical pulse ionization chamber (Hill [ 6 ])

The use of a sheet of organic material for the outer electrode (the main solid surface presented to the counting volume) minimizes alpha contami­ nation background. The other important source of background counts,' namely radon emanating from the constructional materials of the chamber, can be reduced to a low level by suitable choice of material as, for example, ALPHA RADIÓ ACTIVITY,,IN HUMAN TISSUES 295

stainless steel, and by continuous circulation of the counting gas over cooled charcoal. The instrument will accommodate 1.5 g of sample material and it is possible with it to detect and identify nuclides present at concentrations down to 10'13 с/g, while achieving resolutions down to 150 keV. A full des­ cription of this instrument has been given elsewhere[6]. For certain applications, however, it is necessary to develop an in­ strument with higher resolution, and our colleague, Dr. R.V. Osborne, therefore designed a second instrument incorporating a Frisch grid, using a smaller area of sample (1500 cm? as opposed to 15 000 cm2) and a reso­ lution of approximately 50 keV as opposed to 150keV in the larger instrument. A full description of this instrument has been published elsewhere [7]. Its construction is shown schematically in Fig. 3. The cylindrical gas chamber is, again, constructed of stainless steel and encloses a light cylindrical frame carrying an aluminized "Mylar" source. Concentrically placed is a cylindrical grid of stainless-steel wires. A standard alpha source for energy calibration is mounted on one side of a rotatable strip so that it may be ex­ posed to the counting volume at will. The electronics is of fairly standard design, except the head amplifier, specially constructed for the; purpose and employing a modified version of a circuit published by Cottini (mentioned in [7]). The purity of the gas with respect to electron-philic contaminants and radon is again maintained by continuous circulation over heated metallic calcium and charcoal. - With this instrument it is possible to obtain spectra of line widths down to 50 keV, though the quality of the source is very critical if one is to attain this figure. A particular use to which we have put the instrument is the measurement of plutonium and polonium in atmospheric dust derived from the collecting plates of a large electrostatic precipitator [Î7-]. From this precipitator the source obtained is quite uniform as a very thin film, so that it may be transferred to the spectrometer and analysed directly. Con­ siderable care has, of course, been taken to keep the background counting- rate as low as possible. A good source with 50 keV line width may have a contamination of between 5 and 10X 10“3 dpm/nuclide. In order to be able' to deal most effectively with samples that are essentially weightless, the chamber has been designed in such a way that the complete cylindrical electrode system can be.ëasily removed and replaced with a gridded parallel plate system fitted with a source changer capable of accommodating four 25-cm2 sources which can be selected and counted in turn without opening the chamber. The line widths obtainable with this ar­ rangement from a good source are, again, about 50 keV but due to the smaller size of the counting volume; the background counting-rate is ap­ preciably lower than that of the cylindrical system, being equivalent to a source contamination of between 0.5 and 5X 10"3 dpm/nuclide (fo r a 50 keV line width). The design of this electrode system is illustrated in Fig. 3. A fuller description of the complete instrument and its associated equipment will be given elsewhere [7]. . It is found in practice that the spectroscopic and total alpha counting techniques are complementary; normally samples are first total-counted and then analysed spectroscopically, and the two sets of counting-rates and identifications so obtained are compared. Apart from the value of having two quite independent sets of measurements, this procedure provides 296 W .V . MAYNEORD and С . R. HILL

CYLINDRICAL GRIDDED PULSE IONIZATION CHAMBER

ALTERNATIVE PARALLEL PLATE ELECTRODE SYSTEM

' ' , ' F i g .3 '

Dual-purpose high-resolution alpha pulse ionization chamber (Osborne [14])

a check on the introduction of contamination in the process of spectroscopic source preparation. In addition, the simplicity of the total counting technique makes it possible to survey a much larger number of samples than could be examined in the spectrometers. ALPHA RADIOACTIVITY IN HUMAN TISSUES 297

4. SPECTROSCOPIC RESULTS

,4.1. Foods

The spectroscopic technique may be used for a variety of materials. We show examples of spectra obtained from Western Australian wheat, English white bread, Fig. 4, and threç rather exotic materials: (a) cockles, showing the presence of uranium and polonium-210; (b) the "infamous" B razil nut; (c) the taro tuber - a staple m aterial in the diet of the inhabitants of the Island of Niue in the South Pacific, Fig. 5. We also include the spectrum of the residue obtained by evaporation of a drinking water from a granite region.

« - ENERGY (MeV) AUSTRALIAN WHEAT ASH (TOTAL SPECIFIC «^-ACTIVITY, 33 fi(ic/g ASH)

«-ENERGY (M«V)

ENGLISH WHITE BREAD ASH ( TOTAL SPECIFIC « -ACTIVITY, 0. 55 fip c /g ASH)

• Fig-4

Alpha spectra of wheat and bread ashes 298 W.V. MAYNEORD and C.R. HILL

236 u 400- ,238 RdTh {5.42 MeV) Rn (5 .48 MeV)

i\l\i ■ 2000 ThX (5.68. MeV). \ * 2 0 0 - < RaA (6.00 MeV) Vi ■ x ] ! Tn ( 6.26 MeV)

ThA (6.78 MeV) j |R a C '( 7.68 M eV) 0 1 0 0 0 - 5 6 7 8 9 ThC' (6.78 MeV) U «-ENERGY (MeV) " ALPHA SPECTRUM OF ASHED COCKLES (BANGOR)

■ \ 0 -t— 4 5 6 7 8 9 ■ « - ENERGY (MeV) ALPHA SPECTRUM OF BRAZIL NUT TREE ASH

,,236 I 226 lu 2 0 0 0 - 400 < X <2 X

RaC to 200 - оZ> A / и A T------1-----1------1---- 0 4 5 6 7 89 ¿.567 <*- ENERGY (MeV) <* - ENERGY (MeV) ALPHA SPECTRUM OF A SAMPLE ALPHA SPECTRUM OF TARO OF DRINKING WATER (ST. IV E S ) ‘ TUBER ASH

' ' Fig-5

Alpha spectra of foodstuffs and drinking water

4. 2. N o rm a l human tissues

Alpha spectroscopy has been carried out by us on a number of normal human tissues obtained at post mor.tem from cases of accidental deaths, and deaths from malignant disease. The results, some of which are illustrated in Fig. 6, have been reported in detail elsewhere [8]. Perhaps the most striking feature of these observations is the fact that polonium-210 has been identified in all the tissues examined and in many cases appears to con­ stitute the most important single nuclide present. The spectroscopic measurements are best made on material which has been ashed but this, of course, introduces uncertainty as to what fraction of the polonium present has been lost. Our later measurements of the ALPHA RADIOACTIVITY IN HUMAN TISSUES 299

4 5 6 7 8 « - ENERGY (MeV) «-ENERGY (MeV) NORMAL HUMAN LIVER ASH NORMAL HUMAN KIDNEY ASH (SUBJECTS 12,13 8,14) (SUBJECTS 12,13 8.14)

ш z 2 - Í 100 о i—1Л z Э о о 0 ¡y V \ y j> ' 4 5 6 7 8 9 «-ENERGY (MeV) «-ENERGY (MeV) NORMAL HUMAN BONE ASH NORMAL HUMAN AORTA ASH (SUBJEC T 12) (SUBJECTS 12 ,13 & 14)

ix-ENERGY (MeV) «-ENERGY (MeV) NORMAL HUMAN LUNG ASH NORMAL HUMAN LUNG ASH (SUBJECTS 4 8,5) (SUBJECT 6)

F ig . 6 ^

. Alpha spectra of normal human tissue ashes polonium contents have, therefore, been made by a plating-out technique from solutions derived from the tissues. In many of the soft tissue samples the radium-226 levels have been too low for positive identification, but the results are consistent with values found by HURSH et al. [9] using the ema­ nation technique. It must be remembered that the gas of the spectrometer is continually being circulated and radon extracted very efficiently. It has been frequently observed by us that the intensities of the Rn, RaA and RaC' peaks are approximately one half of that of the radium-226, the balance presumably being due to the fraction of the radon escaping from the source. 3 00 W.V. MAYNEORD and C.R. HILL

These factors need to be borne in mind in the quantitative interpretation of the spectra. Even if the radium peak is .small, it does not follow that the alpha dose received by the tissues from radium and its daughter products is necessarily less than that from polonium-210. It is to be noted that lungs are exceptional in that polonium is not the predominant single nuclide but that they accumulate measurable amounts of both thorium-232 and uranium, and also possibly thorium-230. Lung is, indeed, the only human tissue in., which we have observed thorium-232, the obvious suggestion being that the lungs retain relatively insoluble particulate material. One of our spectra, Fig. 6, appears to indicate the presence in the lung of plutonium-239, but further more conclusive evidence of this matter is ' presented later. On the evidence of fluorimetric measurements, natural uranium has been reported to occur in the kidneys of unexposed subjects at concentrations of the order of 2 цg (which is the equivalent of 1.4 pc) per 100 g tissue [10], but the present results indicate uranium concentrations in the kidney lower by a factor of at least 30.

5. TECHNIQUE FOR SPECIFIC ESTIMATION OF POLONIUM-210

While the spectroscopic and total alpha counting techniques give one an excellent overall picture of the activity present, it does not follow that they are necessarily the best means for studying specific nuclides. This has been particularly obvious to us in relation to polonium and its various isotopes, particularly polonium-210. Polonium is normally a very volatile material and a technique involving wet digestion and subsequent plating of the polonium directly onto silver from solution has great advantages in concentrating activity from an initial large mass and avoiding losses in ashing [11]. In a sim ilar way, for the measurement of radium-226, a method such as that used by LUCAS [12] depending upon the release of radon from fairly large quantities of materials, caniead to a technique of high sensitivity and high specificity. It is apparent from the spectra that the main component of alpha activity in human soft tissues tends to be polonium-210 and it is therefore of interest n to record values of concentration of polonium-210 in soft tissues as obtained by a technique specific to that element. . . The technique employed is a standard one [11] depending on the selective deposition of polonium on silver, and we have used it to carry out a.number of measurements on polonium in the human environment as, for example, in the atmosphere, in food and in water. We will, however, leave these results aside for the moment and in this paper report only measurements of polonium-210 in normal human tissues.

6. MEASUREMENTS OF POLONIUM-210 CONCENTRATIONS IN HUMAN TISSUES .

A number of publications report the concentration of polonium-210 in the skeleton of normal human beings [13], but its presence in the human body is, of course, not confined to bone to the extent to which this is the case for ALPHA RADIOACTIVITY IN HUMAN TISSUES 301

TA BLE I

POLONIUM-210 CONTENT OF VARIOUS NORMAL HUMAN TISSUES

N o. o f P o 210 a a c tiv ity ( p c / 1 0 0 g w et) Meandose-rate Tissue sam p les (range of values) (m e a n ) (mrem/yr*) .

Bone 6 0 . 5 - 3 . 1 1 . 7 16

Bone (Eskimo) 1 - • 7 5 . 0 7 4 5

Liver 9 0 . 7 - 1 . 8 1 . 0 1 0

K idn ey 2 0 . 5 - 0 . 9 0 . 7 7

Spleen 3 0 . 1 - 0 . 6 0 . 3 3

Lung 4 0 . 1 - 0 . 4 0 . 3 3.

M u scle 6 0 . 0 3 - 0 . 4 0 . 1 3 1 . 3

T estis 7 . 0 . 0 5 - 0 . 5 0 . 2 4 2 . 3

O vary 1 - 0 . 7 7

P an creas 1 - 0 .3 3 t H air 2 4 . 5 , 1 0 . 9 7 . 7

* Taking the energy of the Po 210 alpha particle as 5.3 MeV, and assuming a relative bio- ■ logical efficiency (RBE) of 10. radium. Appreciable amounts resulting in appreciable dose-rates occur in other tissues, including gonads. Table I shows the polonium-210 content as.measured by us fo r various normal’ human tissues and includes data given by OSBORNE [14]. ' The high level in the Eskimo bone is of interest and might be a reflec­ tion of high meat diet coupled with a high uptake of polonium-210 or lead-210 by animals included in his food. We merely remark at this stage thát if the energy of the polonium-210 alpha particle is 5.3 MeV, and we assume a "relative biological efficiency" of 10, we obtain the mean dose-rates in mrem/yr given in the last column of the Table. It will be seen that even in the absence of such special circumstances as the Eskimo concentration the tissue dose-rates due to polonium-210 could constitute an appreciable fraction of total natural background to those tissues. The high values in the kidneys and gonads are to be noted and emphasize again that polonium is certainly not a bone-seeker, though the biochemistry of its natural dis­ tribution in the body seems to be almost entirely unknown. It is perhaps interesting to remember that some of the earliest autoradiographs ever taken (by Lacassagne and Lattes in 1924 [15] for the tissues of a rabbit) show very clearly the concentration of polonium-210 in the kidneys and pla­ centa. Both the m icro- and macro-distributions of polonium in the human body are worthy of much more study. ; 302 W.V. MAYNEORD and C.R. HILL

We do not wish to discuss here in detail the origin of this polonium-210 which eventually finds itself in the human body. Much is clearly ultimately derived from the radon diffusing from the soil into the atmosphere and the subsequent deposition of daughter products. However, the modes of entry into the human body of chief significance appear to be inhalation of long- lived activity and ingestion in food. It seems probable that in normal human beings the decay of radium-226 in the skeleton provides only a few per cent of the polonium burden present at any particular time. Little correlation is therefore to be expected between body burdens of radium-226 and polonium-210 in any individual [13]. This is confirmed by our observations on the polonium-210 levels in the tissues of those having abnormal radium- 226 burdens. We have made measurements on eight such individuals, whose detailed measurements will be reported elsewhere. In bone and liver of normal individuals polonium-210 appears to be in equilibrium with its parent lead-210. In kidneys and ovaries, however, the polonium is significantly in excess of such equilibrium and is presumably, therefore, largely present as a result of direct uptake of polonium-210 from the blood. It is interesting to compare the absolute concentrations deduced spectros­ copically with those subsequently obtained by the plating technique. Although the comparison can only be approximate, it may be noted that the mean value observed in the liv e r samples for polonium-210 on the basis of the spectros­ copy would be approximately 0.5 pc/100 g wet tissue, whereas the value obtained by the plating technique is approximately 1.0 pc/l00g. The values observed spectroscopically for kidneys are very variable but would seem tô indicate a mean value of the order of a half that obtained by the. plating technique. We believe this does reflect the loss of polonium-210 on ashing prior to spectroscopy. The measurements of particulate activity in the lungs take on a special significance in relation to possible hazards arising from smoking or atmos­ pheric pollution.

7. CONCENTRATIONS IN EXPOSED PERSONS

7.1. Tissues of radium workers

The spectra discussed so far have been those of normal subjects but the development of these spectroscopic techniques also provides the oppor­ tunity to study the metabolism of alpha emitters in abnormal circumstances. As an example we may discuss the case of a man employed between 1909 and 1915 as a manager of a radium refinery in London, England, where he was presumably exposed to inhalation and possible ingestion of uranium ore dust, radium and radon. He subsequently developed chronic lymphatic leukaemia and died in 1958 of a coronary thrombosis. His total bone activity was about 100 times above normal with soft tissue activities only some three or four times above normal. Alpha spectra of liver, kidney, bone and lung are il­ lustrated in Fig. 7. It is of interest that radium-226 is barely detectable in either liver or kidney, although it evidently accounts for most of the ab- ALPHA RADIOACTIVITY IN HUMAN TISSUES 303

b-1Л z

«-ENERGY (MeV) KIDNEY ASH FROM A RADIUM WORKER (SUBJECT 16)

Po2,° ,

750 - «-ENERGY (MeV) LUNG ASH FROM A RADIUM WORKER 234 (SUBJECT 16) . ш 500 - u23e,U ,

Rn RaA R aC ' I I I ThX Th ThA ThC' 250 . ' 1 1 Г ,! U jy ’-V \/\ .. 0 0-Ï1 Ac 227 THEORETICAL u Paf^j_ DAUGHTERS ACTINIUM SERIES SPECTRUM

4 5 6 7 8 9 « - ENERGY (MeV) LIVER ASH FROM A RADIUM WORKER (SUBJECT 16)

o< - ENERGY (MeV) .

BONE ASH FROM A RADIUM WORKER LUNG ASH FROM A URANIUM MINER (SUBJECT 16) - . (SUBJECT 17)

F i g . 7

Alpha spectra of tissues of two occupationally-exposed individuals normal activity of the bone. The somewhat abnormal level of polonium-2lOíí in the liver probably- originates from the decay of skeletal radium and máy therefore provide a measure of the extent to which normal tissue polonium levels derive from this source. The apparent presence of actinium-series nuclides in the liver and thorium-230 in the kidney is consistent with available in­ formation on the metabolic characteristics of these materials, if it can be 304 W. V . MAYNEORD and C . R. HILL

KIDNEY - N.J.R.R.R CASE No. 5043 LIVER — N.J.R.R.P. CASE No. 5043

F ig . 8

Alpha spectra of tissues of a New Jersey radium-refinery worker assùmed that a quantity of ore associated with about 100 mg of uranium was inhaled by the subject during the period of his exposure. This presence of actinium nuclides in the liver and thorium 230 in the kidney has also been observed by us in a man who worked as a foreman in a radium extraction plant in New Jersey, Fig. 8. The presence of the actinium-series nuclides is confirmed in both of these cases by the observation of fast pairs of alpha particles due to actinon and actinium A. It will be seen that the alpha spectroscopic technique sometimes enables information to be obtained directly on humans to supplement the animal experimentation evidence on which we so frequently have to rely.

7. 2. T issu es o f a uranium m in e r

We have also examined lung tissue removed surgically from a uranium miner who had developed bronchogenic sarcoma. The spectrum, Fig. 7, indicates that all members of the uranium series are present to some extent together with a small amount of thorium-series nuclides.' There appears to be disequilibrium in the sample between the early long-lived members of the uranium series. In 100-g tissue uranium-238 and 234 each contribute about 9 pc, thorium-230 about 21 pc, and radium-226 about 5 pc to the total alpha activity of the sample. Since these long-lived nuclides might be ex­ pected to occur at or close to equilibrium in uranium ore, it appears that the disequilibrium observed in the lung sample may be due to selective re­ moval of radium or, to a lesser extent, of uranium from dust deposited in the lungs. A similar situation may hold for the lungs of the English radium- refinery worker, though the specific activity and conseqúently the accuracy of the measurement is less in this case. These examples perhaps illustrate how materials entering the lungs together may suffer different fates de­ pending upon their physical and chemical characteristics. ALPHA RADIOACTIVITY IN HUMAN TISSUES 305

8. ATMOSPHERIC AND RESULTING CONTAMINATION

8. 1. A tm o s p h e ric contam ination

The use of high sensitivity alpha-ray spectroscopy in relation to human metabolic data may also be illustrated by work carried out by our colleague, Dr. Osborne, with respect to plutonium originally in the atmosphere. In parallel to the measurements of polonium in tissues we have carried out periodic examinations of the concentrations of both polonium-210 and plutonium-239 in the atmosphere, using the electrostatic precipitation technique referred to above [16, 17J. We have also measured plutonium-239 in human lungs and other tissues [16]. In view of the extremely low specific activities likely to occur in tissues, and the serious possibilities of confusion with other alpha emitters, particularly polonium-210, we have adopted a procedure whereby the tissue sample is initially "spiked" with plutonium-238, and plutonium then chemically extracted from the sample and assayed using the parallel plate spectrometer. Any plutonium-239 present is determined, and the chemical yield found from the counts in the plutonium-238 peak and the effectiveness of the chemical action discriminating against polonium-210 are thus checked. The spectrum as obtained by Osborne, using the sm all source facility in the spectrometer, is shown in Fig. 9. It is also of interest to show the alpha spectra of atmospheric dust col­ lected during O ctober 1962 and M arch 1963, showing the re la tive amounts of plutonium and polonium at different times following nuclear tests, Fig. 10. It is interesting that the total alpha activity as obtained from the zinc sul­ phide scintillation technique, or as calculated from the alpha spectrum, is the same to within about 10%.

. F i g . 9

Alpha spectrum of plutonium extract from human lungs (spiked with Pu238). (31.3-h count). 306 W.V. MAYNEORD and C.R. HILL

л -ENERGY (MeV)

Fig-10

Typical spectra of long-lived alpha activity in atmospheric dust at Sutton, Surrey

TA BLE II

PLUTONIUM-239 IN HUMAN LUNGS, SPRING 1962 (Osborne [16])

Average lung weight Plutonium-239 Plutoniuin-239

(pooled samples) ( 1 0 ' 15 c / g ) average per person (g ) (p c )

1 1 8 0 ' 0 . 1 2 ' 0 . 1 5

790 0 . 2 1 0 . 1 7 .

1 1 1 0 0 . 1 3 0 . 1 5

8. 2. Contamination o f human lungs .

The absolute levels of plutonium-239 observed in normal human lungs during Spring 1962 are shown in Table II. The concentration of plutonium-239 in the lungs is approximately 0.1 to 0.2X 10~i5c/g. These levels are, in general, lower than those reported for specimens taken during 1954 to 1959. The average lung burden of plutonium-239 found by Osborne (0.16 pc) was approximately 6% of the average polonium-210 activity measured in other individuals. The average total alpha activity of adult lungs in Great Britain has been given as 8 pc. The results of spectroscopic assays following che­ mical separation are in approximate agreement with those previously ob­ tained, using direct spectroscopy without chemical separation [8]. Osborne's results are in good agreement with the lowest values for lungs reported by ALPHA RADIOACTIVITY IN HUMAN TISSUES 307

KREY et al. [18] of Isotopes Incorporated, but we have failed to find any values comparable with those some twenty times higher,, also reported by the same authors. . In this paper we do not wish to discuss the-^complex dosimetry of polonium-210.in the body, nor the undoubted contribution made by a wide range of other polonium isotopes. Our aim is rather to illustrate techniques we have found useful in the study of alpha emitters of interest in human .tissues. In general, it w ill be found that a. combination of techniques is ne­ cessary and that even alpha spectroscopic, techniques are more powerful and useful when combined with simple total alpha counting.

. - REFERENCES, ■

[1]' TURNER, R .C ., RADLEY, J.M . and MAYNEORD, W. V !, Brit. J. Radiol. 31 (1958) 397. [2] . TURNER, R.C., RADLEY, I. M. arid MAYNEORD, W. V ., Nature 181 (1958) 518. ; [3] MAYNEORD, W. V ., RADLEY, J.M . and TURNER, R .C ., Stiahlentherapie 110 (1959) 431. [4] TURNER, R.C., RADLEY, J.M. and MAYNEORD, W.V., Hlth-.Phys. 1 (1958)" 268. ;■ [5] TURNER, R.C., RADLEY, J.M. and MAYNEORD, W. V ., Nature 189 (1961) 348. . .■

[ 6 ] HILL, C.R., Nucl. Inseum. Meth. 12 (1961) 299. ., [7] OSBORNE, R. V.'and HILL, C.R. r Nucl. Instrum. Meth. (in press). .

[8 ] HILL, C .R ., Hlth Phys. 8 (1962) 17. ’ ' [9] HURSH, J.B.' and LOVAAS, A., Nature 198 (1963) 265.’ ’ ' [10] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Report of Committee 2 (1959), ■ Table 12, Hlth Phys. 3 (1960). ' ' ' , [11] BLACK, S.C. , Hlth Phys. 7 (1961) 87. •. . . ' ' , [12] LUCAS, K.F., Rev. sci. Instrum. 28 (1957) 680. . . [13] HOLTZMAN,. R. B ., Hlth Phys., 9 (1963) 385...... [14] OSBORNE, R.V., Nature 199 (1963) 295! . . . . . [15] LACASSAGNE, A., and LATTES, J., C.R. Acad. Sci. Paris Г78 (1924)'488.' ’ [16] OSBORNE, ' R. V.. Nature 199 (1963) 143. ’ . ’ [17] PARKER, R.P. and HILL, C .R ., "Sampling for low concentrations of airborne activity", Environmental Monitoring, Pergamon Press (in press). - • • • ' [18] KREY, P.W. BO GEN, D. and FRENCH, E., Nature 195 (1962) 263. '

DISCUSSION" . . ' .

G. W. DOLPHIN: . Do, you ash the respiratory lymph nodes with the lung m aterial, or do you take out the respiratory lymph nodes separately? I think there could be some interesting differences between the lymph node contents and the m aterial in the main part of the lung. ‘ W. V. MAYNEORD: Usually we do not measure lymph nodes separately, but we have done so in a few cases and have not found any appreciable dif­ ferences from the rest of the lung. This appears also to be true when one is measuring Po210 separately, but,we have few measurements. . W. STAHLHOFEN: First, I would,like to comment briefly on the high Po210 activity in Eskimo bone. We have measured thé Po210 activity in reindeer bone and found an activity of 5 x 'l O " 12 с/g of bone. This means that the activity in soft liver and kidney tissue is of the order of lX10_12c/g of .soft tissue. I agree with P rofessor Mayneo.rd that consumption of reindeer 308 W.V. MAYNEORD and С. R. HILL meat by Eskimos is one of the reasons for the high polonium activity in their bones. We have also studied the normal Po210 content of the human body and have found comparable results. • I would like to make a further comment. I have developed a method for meásuring total alpha-activity similar to Professor Mayneord's. The main difference is that I prepared a homogeneous mixture of the material to be measured and the ZnS scintillator. In this way I obtained a higher counting efficiency. This method is suitable for samples which are of white colour in the ash. I shall describe this method in my paper in these proceedings on measurement of the natural content of Th228 and Ra226 and its daughters in the human body. W. V. MAYNEORD: We are very interested in Dr. Stahlhofen's tech­ nique but were a little afraid of the effects due to colour, texture and particle size of the material to be measured. N. A. TAYLOR: I would like to ask Professor Mayneord if he has been able to derive an effective half-life of Po210 in the body from the balance between the intake of Po210 and the body, content, which is presumably in equilibrium, and if he has examined excreta from normal persons for Po210. It would be interesting if the balance between intake and body content confirmed the effective half-life in the body of about 36 d which I and others have observed by excretion analysis following accidental intake of Po210. W.V. MAYNEORD: We have not-derived an effective half-life in this way, partly because of lack of sufficiently accurate data on food intake. So far as faecal excretion is concerned, as you probably know, we did measure total alpha-activity, but we have not yet made specific measure­ ments o f the polonium only. I hope we will be able to do so. I agree, of course, that this would be a very valuable'way of supplementing your data. J.K. MIETTINEN: I should like to point out that the measurement of the biological half-life of Po210 solely on the basis of intake through food and excreta analysis may lead to grossly wrong results if the Pb210 content of the bones is not also reckoned with. • W.V. MAYNEORD: This is a good point. We agree that both P o and Pb should be considered in this context. W. JACOBI: Can you estimate, on the basis of the measured natural Po210 content of different human tissues, the fraction due to inhalation of Rn222 and- its decay products in a ir and that due to intake- through food and drinking water? _ . . W.V. MAYNEORD: I don't think so. Certainly, inhalation is an im ­ portant method of entry for Po210, but food - I would think - was the pre­ dominant one, particularly in the cáse of Eskimos. It will obviously depend on the kind of diet. Holtzman has made such an estimation for Pb210 (Health Phys. 9 (1963)385). ' • . . P. KAYSER: I would like to ask whether you first áttempted to separate chemically the alpha emitters from the inert substances before submitting the alpha emitters to spectroscopy. ' W. V. MAYNEORD: In the case of the Osborn chamber this has been the normal procedure; the sm all source arrangement of the Osborn chamber is intended for use following chemical separation of a specific element. In the case of our original experiments we were, of course, concerned with all the materials present and we therefore mérely divided and looked at the. ALPHA RADIOACTIVITY'IN HUMAN TISSUES 309 whole m aterial, with no chemical separation. Indeed, one of the reasons for developing this technique was that we, at one stage, became very suspicious of the alpha activities which are to be found in most chemical reagents. And I would advise anyone who enters this field to have a good chemist who knows how to look for alpha particles in his reagents. . -

THE USE OF GAMMA-RAY SCINTILLATION SPECTROMETRY IN BIOASSAY

■ A. HOLMES HEALTH PHYSICS AND MEDICAL DIVISION, . ATOMIC ENERGY RESEARCH ESTABLISHMENT, HARWELL, ENGLAND (PRESENTED BY A. MORGAN)

Abstract — Résumé — Аннотация — Resumen

THE USE OF GAMMA-RAY SCINTILLATION SPECTROMETRY IN BIOASSAY. The use of gamma-ray scintillation spectrometry for the determination of radionuclides in bioassay is discussed. This technique has the advantage that direct measurements can be made on untreated samples, thus avoiding tedious and time­ consuming radiochemical separations. A second advantage is that a number of radionuclides may be deter­ mined simultaneously in the same sample. The equipment used for gamma- ray spectrometry at the Atomic Energy Research Establishment (AERE), Harwell, is described. Data are recorded automatically on punched cards and a computer programme is used, to deduct the background and to plot the corrected spectrum, which is suitable both for measurement and record purposes. The same programme can be used to deduce the relative amounts of a number of known radionuclides in a sample, by a least squares fitting method. The detection limits for this equipment are calculated for five gamma emitters with energies in the range 0.134 to 1.52 MeV and for counting periods of 1, 10, 100 and 1000 min. Efficiency factors are also calculated for various sample volumes from 400 to

2 0 0 0 m l. The sensitivity of this method for a number of radionuclides commonly encountered in the atomic energy industry is defined in terms of ‘action* levels in excreta. These 'action* levels are derived from maximum permissible body burdens, recommended by the International Commission on Radiological Protection (ICRP), and an assumed conservative excretion rate of 0 . 1 % per day of the maximum permissible body burden (mpbb). They are expressed in picocuries of gamma activity, to take the branching ratio into account. Almost all the radionuclides considered can be detected at one tenth or less of the ’action' level, with a ten-minute counting period. A number of examples of gamma-ray spectrometric determinations of contaminated urine and faecal samples from occupationally-exposed personnel are also included.

EMPLOI DE LA SPECTROMÉTRIE GAMMA DANS LES ANALYSES BIOLOGIQUES. Le mémoire est con­ sacré à l'emploi d'un spectromètre à scintillation gamma pour la détermination des radionucléides dans les échantillons soumis à des analyses biologiques. La méthode présente l’avantage de permettre des mesures directes sur des échantillons n’ayant fait l'objet d'aucune préparation,, ce qui évite des séparations radio- chimiques longués et fastidieuses. Un deuxième avantage est qu'elle permet.de déterminer simultanément plusieurs radionucléides dans un même échantillon. ' L'auteur décrit l’équipement qué le Centre de Harwell utilise pour la spectrométrie gamma. Les données sont enregistrées automatiquement sur cartes perforées; un programme de calculatrice permet de déduire le bruit de fond et de tracer la courbe du spectre corrigée, qui se prête aussi bien à la mesure qu’à l’enregistre­ ment. On peut utiliser le même programme pour déduire les quantités relatives d'un certain nombre de radionucléides connus dans un échantillon, en opérant un ajustement par la méthode des moindres carrés. Les limites de détection de cet appareillage ont été calculées pour cinq émetteurs gamma ayant des énergies comprises entre 0,134 et 1,52 MeV et pour des périodes de comptage de 1, 10, 100 et 1000 min. On a également calculé les facteurs d’efficacité pour divers volumes d’échantillons allant de 400 à 2000 cm 2. La sensibilité de cette méthode, dans son application à un certain nombre de radionucléides que l'on rencontre couramment dans l’industrie atomique, est définie en «niveaux d’action» dans les ¿xcreta. Ces «niveaux d’action » sont établis à partir des charges corporelles maximums admissibles (CCMA) recomman­ dées par la CIPR, et d’un taux d'excrétion supposé qui a été évalué d’une manière prudente à 0,1% par jour de la CCMA. Ils s'expriment en picocuries d’activité gamma, de manière à tenir compte du rapport d’embranchement. A peu près tous les radionucléides étudiés peuvent être décelés à un dixième du «niveau d’action » ou moins, avec une période de comptage de 10 min. Les mémoires donnent également un certain

311 312 A. HOLMES nombre d'exemples de détermination par spectrométrie gamma de spécimens d'urines et de matières fécales contaminées, provenant d*un personnel professionnellement exposé.

ИСПОЛЬЗОВАНИЕ СЦИНТИЛЛЯЦИОННОЙ ГАММА-СПЕКТРОМЕТРИИ В БИОЛОГИ­ ЧЕСКИХ ИССЛЕДОВАНИЯХ. Обсуждается использование сцинтилляционной гамма-спектро­ метрии для определения содержания радиоизотопов в биологических образцах. Преимуществом метода является прямое измерение необработанных образцов без предварительного радио­ химического разделения, требующего значительной затраты времени. Вторым преимуществом является возможность одновременного определения нескольких радиоизотопов в одном и том же образце. . - ( Дается описание оборудования, используемого для гамма-спектрометрии в Научно­ исследовательском центре по атомной энергии в Харуэлле. Данные автоматически регистри­ руются на перфорированных карточках, используется вычислительная программа для вычита­ ния фона и вычерчивания корригированного спектра, который можно использовать как для измерения, так и для регистрации. Эту же программу можно использовать для прослежи­ вания относительных количеств ряда известных радиоизотопов в образце путем аппроксимапии способом наименьших квадратов. Пределы обнаружения для этого оборудования рассчитаны для гамма-излучателей с энергиями от 0 ,134 до 1,52М эв и для периодов счета 1; 10; 100 и 1000 минут. Рассчитан также коэффициент эффективности при объеме образцов от 400 до

2 0 0 0 м л . • Чувствительность этого метода для ряда радиоизотопов, используемых обычно в атомной энергетической промышленности, определяется в зависимости от "действительных" уровней в выделениях. Эти "действительные" уровни определяют на основе максимальных допустимых величин содержания изотопов в организме, рекомендуемых МКРЗ, и константы (в большинстве случаев консервативной) скорости выделения порядка 0 , 1 % в день от максимально допустимой величины содержания в организме. Их обозначают в пикокюри гамма-активности, чтобы учитывать коэффициент ветвления. Почти все рассматриваемые радиоизотопы могут быть обнаружены в дозах от одной десятой или менее от "действительного" уровня в течение десятими­ нутного периода подсчета. Приводится несколько примеров гамма-спектрометрического исследования содержащих изотопы образцов мочи и экскрементов рабочих, подвергающихся профессиональному облучению.

APLICACIÓN DE LA ESPECTROMETRÍA GAMMA POR CENTELLEO EN LOS ANÁLISIS BIOLÓGICOS. El autor examina la aplicación de la espectrometría gamma por centelleo para determinar radionúclidos en muestras destinadas al análisis biológico. Esta técnica presenta la ventaja de permitir mediciones directas en muestras no tratadas, con lo que se evitan las tediosas y largas separaciones radioquímicas. Otra ventaja es que pueden determinarse simultáneamente varios radionúclidos en la misma muestra. . Se describe el equipo de espectrometría gamma utilizado en el Atomic Energy Research Establishment de Harwell. Los datos se registran automáticamente en fichas perforadas y se utiliza un programa de calcula­ dora para deducir la actividad de fondo y representar el espectro corregido, que puede utilizarse con fines de medición y de registro. El mismo programa puede aprovecharse para determinar las cantidades relativas de un número conocido de radionúclidos presentes en una muestra, aplicando el método de ajuste por cuadrados mínimos. Los límites de detección del equipo se han calculado para cinco emisores gamma de energía com­ prendida entre 0,134 y 1,52 MeV y para tiempos de recuento de 1, 10, 100 y 1000 min. También se han calculado los factores de rendimiento correspondientes a volúmenes de muestra comprendidos entre 400 y

2 0 0 0 m i. ■ La sensibilidad de este método para una serie de radionúclidos que se utilizan corrientemente en la industria de la energía atómica se define en función de los grados de « actividad » de las excreciones. Estos grados de «actividad» se determinan a partir de las cargas corporales máximas admisibles, recomendadas por la Comisión Internacional le Protección Radiológica (CIPR), y de una velocidad hipotética de excreción (fijada según criterios prudenciales) de 0,1 °¡o al dfa de la carga corporal máxima admisible. Se expresan en picocuries de actividad gamma, a fin de tener en cuenta la razón de bifurcación. Casi todos los radio­ núclidos considerados pueden detectarse a un nivel por lo menos diez veces inferior al grado de «actividad», con un tiempo de recuento de 10 min. Se incluye buen número de ejemplos de determinaciones por espectro­ metría gamma de muestras de orina y heces contaminadas, procedentes de personas expuestas profesionalmente. GAMMA-RAY SCINTILLATION SPECTROMETRY IN BIOASSAY 313

1. INTRODUCTION

Increasing amounts of radioactive material are being handled each year in atomic energy establishments, in universities and in industry. To ensure that internal radiation exposure is kept within permissible limits, extensive radiological protection measurements are made on behalf of the people work­ ing with radioactive material.- These measurements include environmental monitoring for air and surface contamination and bioassay determinations on samples of excreta. Internal contamination with radionuclides which emit penetrating X- or gamma radiation can be detected by in vivo counting of exposed personnel. MEHL and RUNDO [1] have calculated typical detection limits for whole-body monitors, and in almost all cases these are very much less than the relevant -maximum permissible body burdens (mpbb) recommended by the ICRP [2]. It is not practicable, however, to use whole-body monitoring for the routine examination of all people potentially exposed to internal contamination. Whole- body monitors are expensive and require to be operated by highly specialized staff, so that generally they are available only in very large establishments. Those which are available are more usually employed to investigate cases of known accidental exposure, or for metabolic studies in which a radio­ nuclide may be introduced deliberately into the body. For reasons of policy and convenience therefore, routine sampling and analysis of excreta is generally used to detect cases of internal contamina­ tion. In a bioassay programme it is desirable to limit, as far as possible, the expenditure of time and effort spent on routine determinations. At AERE, Harwell, gamma-ray spectrometry is used, where applicable, for routine determinations of radionuclides in bioassay samples. By the direct counting of untreated samples, many tedious radiochemical separations are avoided. The equipment used at AERE and its sensitivity for gamma-emitters of dif­ ferent energies in various samples is described in this paper.

2. PRINCIPLES OF GAMMA-RAY SCINTILLATION SPECTROMETRY

It is beyond the scope of this paper to elaborate the principles and prac­ tice of gamma-ray spectrometry, about which much has been written al­ ready [3,4] . It is sufficient to point out that individual gamma rays are de­ tected, and their energy measured and recorded by the spectrometer, in such a manner .that the operator is provided with a m easure of the d is tri­ bution and intensities of gamma-rays from the source of radioactive ma­ terial. Since each gamma-emitting radionuclide has a characteristic range of radiation energies, it is usually possible to identify and measure the nu­ clides in the sample quite simply. Sodium iodide scintillation spectrometers are able to resolve medium energy photons of at least 8-10% difference in energy, providing their inten­ sities are similar. In the presence of interfering activities,'"'the accuracy with which gamma-ray energies or intensities can be measured is limited by this resolving power and by instrument stability. It is usually possible to measure gamma-ray energies to an accuracy of at least 1% and their 314 A; HOLMES intensities to that approaching the best standard source available (i.e. 2 to3%), all these figures being subject to the statistical uncertainties of the counting procedures.

3. EQUIPMENT AND PROCEDURE

The equipment described is that used at the present time in the Bioassay Section at AERE. The detector consists of a cylindrical‘3x3-in thallium- activated sodium iodide crystal.which is mounted on a photomultiplier, oper­ ating in conjunction with a linear amplifier and multichannel puls e-height analyser.. The crystal and photomultiplier are contained in a 4-in thick steel shield. . • ■ Skin and blood samples, nasal swabs and other miscellaneous samples are counted directly on top of the crystal. Urine samples are counted in polythene containers of various capacities shown in Figs. 1 A and B, which are designed to surround the detector with a uniform layer of liquid giving optimum counting geometry. Faecal samples can be counted directly in the containers (see Fig. 1 C) in which they are collected. Usually however, they are ashed, the residue dissolved in a mixture of concentrated nitric and hydrochloric acids, diluted to 200 ml, and counted in the special polythene container shown in Fig. ID, • In the Bioassay Section at AERE, data from the pulse-height analyser are recorded automatically on punched cards. A programme written for either an IBM or 7030 computer is available, which converts the data into counts per minute per channel, subtracts the background and provides a plotted difference spectrum, suitable both for measurement and record pur­ poses. The pulse-height spectrum produced is characteristic of a particular radionuclide or nuclides and the components can often be identified imme­ diately by reference to the appropriate literature [3, 5]. If there is doubt about the identification, confirmation can frequently be obtained by decay measurements on the sample. In some instances, resort must be made to chemical separations, but in such cases the main advantage of spectrometry has been defeated. The gamma-ray energies and intensities corresponding to the total absorption peaks (photopeaks) of the observed pulse-height spec­ trum, are determined by means of standards of gamma-emitting nuclides of known energy and activity. Use can also be made of previously deter­ mined efficiency factors [4]. The intensity of the gamma-ray is measured by the area or height of the relevant photopeak. If the radionuclides present in the sample can be identified unequivocally, a least squares method of analysis can be used for the simultaneous estimation of up to ten components, the calculation being performed by a computer. This method [6] is superior to the usual graphical methods of quantitative interpretation.

4. DETECTION LIMITS FOR GAMMA-EMITTING NUCLIDES

i For gamma-emitters, the limit of detection for a particular detector depends upon a number of factors, including gamma-ray energy, sample volume and geometry and the time for which the sample is counted. The GAMMA-RAY SCINTILLATION SPECTROMETRY IN BIOASSAY 315

(a) (b) SCREW CAP

SCREW CAP

F ig . 1 A F ig . 1 В

400-ml urine container 1 0 0 0 , 1600 and 2 0 0 0 -m l urine container

SAMPLE CON- : ( c ) TAINED IN PLYO- (d)

SCREW CAP

WAXED CARDBOARD

- F ig . 1 C Fig. 1 D

Faecal collection tub 2 0 0 ml oxidized faecal container effect of these parameters on the detection limit of the equipment described in section 3 was studied and the results are outlined below. '

4.1. Effect of gamma-ray energy and counting time . .

According to the ICRP [2] , the standard man excretes 1400 ml of urine daily. Experience at ,AERE has shown that, in practice, the daily urinary output can va ry between 400 and at least 2000 m l. F o r a constant amount of activity, counting efficiency will decrease with increasing sample volume and th erefore 2000 m l was taken as representin g an unfavourable case. 316 A. HOLMES

Using this volume of liquid in the container shown in Fig. 1 B, detection li­ mits* in pc (gamma) were determined over a range of energies from 0.13 to 1.52 MeV using counting periods of 1, 10, 100 and 1000 min. These measure­ ments were made using standard solutions of Ce144, I 131, Cs131, Zn65 and К 42, with estimated accuracies of ±5%, supplied by the Radiochemical Centre, Amersham. The detection limits obtained in this way are tabulated in Table I and va ry from 5 to 390 pc depending on the energy and counting time selected. •

4. 2. E ffe c t o f sam ple volum e

To compare the detection limits for samples of equal activity contained in different volumes, efficiency factors (cpm/pc gamma) were measured for 400, 1000, 1600 and 2000-ml samples in the polythene containers shown in Figs. 1 A and IB . Efficiency factors for point sources and for 200 ml of solution in the polythene container for oxidized faecal samples (see Fig. ID ) were also determined using the same range of gamma-ray energies as be­ fore. To determine the efficiency, factors, the areas of the relevant photo­ peaks were estimated above a base line drawn from the trough on the low- energy side of the peak to the tail at the high-energy side. While this method of approximation is satisfactory for the monoenergetic gamma-emitters (Cs!3i, Zn65 and K42) the baseline cannot be positioned with the same degree of certainty in the more complex spectra of C e144 and I 132. The efficiency factors for these two radionuclides will therefore be subject to a greater error. The efficiency factors obtained for the various configurations at different energies are given in Table II. The figures in brackets represent the efficiency relative to the 2000-ml sample volume.

5. DESIRABLE LIMITS OF DETECTION FOR GAMMA-EMITTING ' NUCLIDES

For gamma-ray spectrometry to be used successfully in a bioassay programme, it must be able to detect radionuclides at concentrations which will be found in urine or faeces from people with burdens amounting to a significant fraction of the mpbb recommended by the IC RP. It is customary in the field of bioassay to refer to the 'action' level for a particular radio­ nuclide. This may be defined as the rate of excretion which might be ex­ pected (on the basis of human experimentation for example) from a person who, has acquired a maximum permissible burden of a radionuclide some time before sampling. The period between acquisition and sampling is nor­ mally set with regard to the biológical and physical half-life of the radio­ nuclide in question. .The rate of excretion and the way in which it diminishes with time differ widely from one element to another and are affected by many factors, in­ cluding the mode of intake, the physical and chemical properties of the con­ tamination and personal Variations. The human experimental data required

' The detection limit is defined as the activity which is equivalent to three times the statistical standard deviation on the background in the region of the relevant photopeak. AM-A SITLAIN PCRMTY N IASY 7 1 3 BIOASSAY IN SPECTROMETRY SCINTILLATION GAMMA-RAY

TABLE I

DETECTION LIMITS FOR SEVERAL GAMMA-EMITTERS IN 2000 ml OF URINE

/»• Gamma energies Background in E ffic ie n cy Limit of detection in pc (gamma) for various counting times . region o f photop eak

(M e V ) (c p m ) (cpm/pc gamma) 1 0 0 0 m in 1 0 0 m in 1 0 m in 1 m in

0 . 1 3 4 ' 9 0 .8 0. 167 5 17 5 4 1 7 2

' 0 . 3 6 7 0 . 9 0 . 0 8 6 ' 9 . . 29 9 2 2 9 3

0 . 6 6 ' 6 2 .4 0. 0 6 9 L 1 1 . 3 4 1 0 9 3 4 4

1 . 1 1 3 2 . 6 0. 0 4 4 13 3 9 126 3 9 2

. 1. 52 1 0 . 6 0 . 0 2 5 1 2 3 9 1 2 4 3 9 2 318 A. HOLMES

TABLE II .

EFFICIENCIES FOR VARIOUS COUNTING GEOMETRIES AND GAMMA-RAY ENERGIES

E ffic ie n cy (cpm/pc gamma)

Energy Point source Sample volume (and container) (M eV ) on cry sta l

2 0 0 m l 4 0 0 m l 1 0 0 0 m l 1 6 0 0 m l 2 0 0 0 m l (F ig . I D ) (F ig . 1А У (F ig . I B ) (F ig . I B ) (F ig . I B )

0. 13 0. 78 0 . 2 4 0. 3 5 0. 23 • 0. 19 0 . 17

( 4 . 6 ) ( 1 . 4 ) ( 2 . 1 ) ( 1 .4 ) ' ( 1 . 1 ) ( 1 . 0 )

0 . 3 6 0 . 4 2 0. 14 0 . 18 ' 0 . 13 0 . 1 0 0 . 0 8 6

(4 . 9) ( 1 . 6 ) ( 2 . 1 ) ( 1 .5 ) ( 1 . 2 ) ( 1 . 0 )

0 . 6 6 0 . 2 9 0 . 1 0 0 . 14 0 . 1 0 0 . 0 8 2 0 . 0 6 9

(4 . 2 ) ( 1 . 4 ) ( 2 . 0 ) ( 1 . 4 ) ( 1 . 2 ) ( 1 . 0 )

1 . 1 1 0 . 19 0 . 0 6 5 0 . 0 8 6 0. 0 6 7 . 0 . 0 5 2 0. 0 4 4

( 4 . 3 ) (1 . 5 ) ( 2 . 0 ) ( 1 .5 ) ( 1 - 2 ) ( 1 . 0 )

1. 5 2 0. 0 9 7 0 . 0 3 5 0. 0 4 6 0. 0 3 5 0 . 0 2 8 0 . 0 2 5

(3 . 9 ) ( 1 . 4 ) ( 1 . 8 ) ( 1 . 4 ) ( 1 . 1 ) ( 1 . 0 )

to make realistic assessments of 'action' levels are only available for a few elements at the present time and therefore a generalized treatment is adopted for the calculation of the 'action' levels used in this report. For most radionuclides excretion in the 24 h immediately after intake (corrected for physical decay) exceeds 1% of the dose, however it is administered. It seems reasonably conservative, therefore, to take a valúe of 0.1% as a rate of excretion at some arbitrary time after intake. Applying this excretion rate to the mpbb recommended by the ICRP, 'action' levels for a number of commonly encountered gamma-emitting nuclides have been calculated and are given in Table III. These levels are expressed in pc gamma and take into account the branching ratios (i.e. the number of photons per disinte­ gration). In Table IV the detection limits of the radionuclides listed in Table III are expressed as a fraction of the 'action'"level and for the m ajori­ ty, the detection limit is below one tenth of the 'action' level, indicating that the sensitivity is satisfactory for bioassay determinations. GAMMA-RAY SCINTILLATION SPECTROMETRY IN BIOASSAY 319

TA BLE III

'ACTION' LEVELS FOR SELECTED GAMMA-EMITTING NUCLIDES

Data for most abundant gam m a-ray N u clid e P h y sical 'A c tio n ' E nergy ■ B ran ch in g 'A c tio n ' h a lf -life le v e l ra tio le v e l ( n c / 2 4 h ) (M eV ) m (п су 2 4 h)

N a 22 2 6 1 0 0. 5 1 a 1 7 .8 . Уг 178

N a 24 15 h 7 1. 37 1 0 0 7 . 0

S c 4 6 8 4 d 1 0 , 0 . 8 9 1 0 0 1 0 . 0

C r 51 2 7 . 8 d 8 0 0 0 . 3 2 8 6 4 . 0

M n 54 2 9 1 d 2 0 0 . 8 4 1 0 0 . 2 0 . 0

M n 56 2. 5 8 d 2 0 . 8 5 99 2 . 0

F e 59 4 5 d 2 0 1 . 1 0 56 1 1 . 2

C o 6 0 • 5. 2 7 yr 1 0 1. 17 1 0 0 1 0 . 0

C u 64 1 2 . 8 h • 1 0 0 . 5 1 a 38 3 . 8

Z n 65 2 4 5 d 60 1 . 1 1 ' 4 5 2 7 . 0

A s 7 6 2 6 . 5 h 2 0 . 0 . 5 6 4 5 9 . 0

S e 75 1 2 1 d 90 0 . 2 7 56 5 0 . 4

B r 82 3 6 h 1 0 0 . 78 83 8 . 3

Sr 85 60 d 60 .0 . 5 1 , 1 0 0 60 . 0

R u 103 4 0 d ' 2 0 0 . 5 0 9 0 1 8 . 0

R u 106 1 3 0 . 51 3 0 • 0 . 9 У1

P d 109 13. 6 h 7 0 . 0 9 4 0 . 3

A g u o m 2 5 3 d . 1 0 0 . 6 6 9 4 9. 4

S b 122 2. 7 4 d 2 0 0. 57 6 6 13 . 2

S b 124 60 d 1 0 0 . 60 99 9. 9

C s 134 2. 19 yr 140 0 . 8 0 90 1 2 6 . 0

C s 137 3 0 yr ■ . 2 1 0 0 . 6 6 . 8 2 17 2 . 2

C e 141 3 2 . 5 d 30 0 . 15 4 9 14. 7

C e 144 2 8 5 d 5 0. 13 1 1 . 0 . 6

W 187 2 4 d 30 0 . 69 32 9 . 6

A u 198 2 . 7 d ' 2 0 0 . 4 1 96 ■ 1 9 . 2

A u 199 ’ 3 . 15 d 7 0 0 . 16 ' 4 2 2 9 . 4

H g 203 . ~ 4 7 d 4 • 0 . 28 . 83 .3 .3 .

Pa 233 27 d 4 0 0 . 3 1 ■ ■ 4 2 16. 8

a Annihilation radiation from positron-emitter. 32 0 A. HOLMES

TA BLE IV

DETECTION LIMITS FOR GAMMA-EMITTING NUCLIDES IN 2000 ml SAMPLES COUNTED FOR 10 min

Limit of detection between 10 " 4 - 1 0 ' 3 of 'action' level

Cs134 C s137

Limit of detection between 10 ‘ 3 - 1 0 " 2 of 'action' level

N a22 C r 51 M n54 Z n 65 S e 75 S r 85 Ru103

S b 122 C e 141 Au198 Au199 P a 233

Limit of detection between 10 " 2 - 1 0 " 1 of 'action' level

N a 24 S c 4 6 M n56 F e 59 С о 6 0 С и м As76

Br82 A g uom Sb124 C e 144 W 187 Hg103

Limit of detection between 10 " 1 - 1 0 -0 of 'action' level

Ru106 Pd 109

6. EXAMPLES .

The following examples, illustrating the use of gamma-ray scintillation spectrometry in bioassay determinations, have been selected from cases of internal contamination arising at AERE, Harwell and at the Radiochemical Centre, Amersham. In none of these cases was the maximum permissible body burden exceeded.

6. 1. Case A

. A pressurized-suit worker in a high activity handling building cut his glove and fin ger on the edge of a contaminated steel plate. The conta­ mination consisted of irradiated uranium and thorium particles containing fission products (Zr95-Nb95, Rui03, Cei4i) and Pa233 respectively. The gam m a-ray spectrum of a contamination smear from the steel plate is shown in Fig. 2 A. , No abnormal gamma activity was detected in urine samples taken after the incident, but the gam m a-ray spectra of a faecal sample (Fig. 2 B) showed the presence of all the major gamma-emitters present in the contamination smear. The possibility that this activity was inhaled can almost certainly be ruled out, as the man was working in a pressurized suit at the time of the incident. The presence of the contaminating radionuclides in faeces suggests that some solubilization of the uranium and thorium particles occurred in the body, liberating the associated fission products and P a 233 into the blood GAMMA-RAY SCINTILLATION SPECTROMETRY IN BIOASSAY 321

CHANNEL NUMBER

. F ig . 2 A

Gamma-ray scintillation spectrum of a contamination smear from a steel plate

F ig . 2 В

<.',amma-ray scintillation spectrum of a faecal sample containing fission products and protactinium-233 streám. Subsequent secretion into the gastro-intestinal tract.via’ the bile duct and excretion in faeces appears to be the most likely route of elimination.

6. 2. Case В -

Urine sampling was requested for a chemist starting work with volatile compounds of selenium labelled with Se75. Gamma-ray spectrometric ana­ lysis of urine samples showed them to contain Se75 and also small amounts 322 A. HOLMES of.Ii31. Measurements made over a period of 100 d proved that, while vola­ tile compounds were being handled, the concentration of Se75 in urine re­ mained fairly constant, but declined steadily when this work was discontinued. The gamma-ray spectrum of a urine sample, together with the fitted curve from a least squares analysis and the difference spectrum are shown in Fig. 3, The spectrum of Cs137 was superimposed on that of the sample

F i g .3

Gamma-ray scintillation spectrum of a urine sample containing selenium-75, iodine-131 and superimposed caesium-137 . to facilitate.automatic gain control by the computer using the 0.66 MeV photo­ peak. A well-defined photopeak is superimposed in this way because gain control is not possible using the non-gaussian photopeaks which occur in samples of low activity. The standards used for least squares fitting were K40, Se^s, 1131 and Cs 137. ' . ' ;

6. 3. Case С . . , .

A man working on top of the DIDO high-flux reactor inhaled radioactive dust when a quantity of helium gas was discharged from an experimental rig. Measurements of body radioactivity after the incident showed that the man had inhaled Z n 6& and A g n0m, but the in itial leve ls w ere less than 1% of the relevant mpbb. Although the initial burdens of these radionuclides were so small, the levels of both Zn65 and AgH°m in faeces could be measured by gamma-ray spectrometry for 700 and 300 d respectively. The excretion rates obtained from these, measurements and the retention pattern deduced from ;whole-bódy monitoring enabled, accurate assessments of the biological half-lives of these two radionuclides to be made [7] in this case. Least squares fitting was used in the analysis of gamma-ray spectra of faecal sairiples. A typical spectrum is shown in Fig. 4, together with the fitted and difference spectra. The 1.11 MeV photopeak of Zn65 was used for gain control in the initial samples and the 1.46 MeV photopeak of К 40 in the later samples, which contained only small amounts of Zn65. During this GAMMA-RAY SCINTILLATION SPECTROMETRY IN BIOASSAY 323

Ag’10™

' Fig.4 ' '

Gamma-ray scintillation spectrum of a faecal sample containing zinc-65 and silver-110 m study the amounts of fallout derived Z r9&-Nb35 in faeces were highly variable, but by including a Z r 95-Nb95 standard among the others used in the least squares fitting procedure it was possible to correct for thèse interfering radionuclides.

6.4. Case О , .

A worker handling material in a high activity cell inhaled a small amount of thulium-170. Although no ábnormal activity was detected in urine samples taken after the incident, the gamma-ray spectra of faecal samples showed that Tm170 was the major contaminant, with smaller amounts of Co60 (pos­ sibly Fe59 ) and Zn65. The presence in the spectrum of the characteristic X-ray of ytterbium at 52 keV, in addition to the photopeak at 84 keV (see Fig. 5 A) establishes the presence of Tm1™ . Cobalt-60 was confirmed by the 2.50 M eV sum peak in the spectrum, but this s till does not preclude the presence of F e59, which has a similar spectrum, but no sum peak. Zinc-65 was confirmed by stripping a standard spectrum of Co60 (normalized on the 1.33 MeV photopeak) from that of the sample, as shown in Fig. 5B. The Zn65 photopeak is revealed in the difference spectrum.

7. CONCLUSIONS *

Gamma-ray scintillation spectrometry, using the methods described above, is a useful technique for the analysis of bioassay samples. Radio­ nuclides present in such samples can often be identified and the amount de­ termined in a short time by'direct counting of the untreated material. The sensitivity of the technique is adequate for almost all gamma-emitters. If 324 A. HOLMES

F ig . 5 A

Gamma-ray scintillation spectrum of a faecal sample containing thulium-170

(1.11M«V) (1.17MCV)

. F ig . 5 В

Gamma-ray scintillation spectrum of a faecal sample showing the presence of zinc-65 after stripping a cobalt-60 spectrum a computer is available, the automatic handling of data makes this .technique even more attractive, as much laborious manual calculation can be avoided.

ACKNOWLEDGEMENTS

The author wishes to thank A. Morgan and L. Salmon for help in pre­ paring this paper and Miss J. Williams who carried out some of the experi­ mental work. GAMMA-RAY SCINTILLATION SPECTROMETRY IN BIOASSAY 325

; REFERENCES .

[1] MEHL, J; and RUNDO, J. , Hlth Phys. 9 (1963) 607. [2] RECOMMENDATIONS OF THE INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION. Report of Committee II on permissible dose for internal radiation, Pergamon Press (1959). [3] CROUTHAMEL, C. E. , Applied gamma -ray spectrometry, Pergamon Press (1960). [4] HEATH, R. L. , Scintillation spectrometry gamma-ray spectrum catalogue IDO 16408 (1957). [ 5 ] STROMINGER, D. , HOLLANDER, J. M. and SEABORG, G. T. , Rev. mod. Phys. 30 (1 9 5 8 ) 585.

[ 6 ] . SALMON, L. , AERE-R 3640 (1961). [7] NEWTON, D. and HOLMES, A. To be published. .

. DISCUSSION

N.I. SAX: Have you any data derived from examining the contamination of organs dissected during autopsy or in the operating room? A. MORGAN: We have not used the techniques described in this paper for the examination of organs taken at autopsy. We have analysed human bone samples for plutonium-239, using a technique similar to the urinalysis described in the paper I presented in these proceedings on the determination of alpha-emitting plutonium in urine using a solid-state counter. But this is a different subject. C.W. SILL: How do you obtain or prepare the polyethylene sample containers of the re-entrant or Marinelli type? . And have you had experience of contamination difficulties in using sample containers made of polyethylene? , A. MORGAN: These containers are fabricated by welding a cylindrical insert into a standard polythene container from which the bottom has been rem oved. ! We have had no difficulties with the contamination of polythene bottles in our bioassay work. In connection with tracer experiments using caesium-137 in river water, however, we found this radionuclide to be strongly adsorbed on polythene and even treatment with strong acids failed to remove it. We have studied the losses of plutonium in urine stored in polythene bottles and, as far as we can discover, adsorption does not occur to any significant extent. C.W. SILL: I would like to make two points relating to this subject. Firstly, it has been our experience that polyethylene is an extremely poor choice for sample containers. The material is very porous and the contents of the container diffuse right into the wall. This is not a question of ad­ sorption on the wall, but of diffusion right into the plastic itself. Polystyrene containers are much superior. Secondly, containers of the Marinelli or re-entrant type can be made very conveniently from a 1-gal paint can witha U.S. N o.2 tin can solderedto an ap­ propriate hole punched in the bottom. The No.2 can fits a standard 3inX3 in sodium iodide crystal and the cans, being already tinned, are very easily joined by soldering. K. G. M cN E ILL; In the case of the man with the cut finger, L)r. Morgan, how do you think the activity entered his body — through the wound or through the mouth, when he automatically licked his finger? . Another point. When you strip away spectra, as in Fig. 5B, do you use the spectrum from a free point source or that obtained with a phantom? 326 A. HOLMES

A. MORGAN: In answer to the first question, the man was working in a pressurized suit when he cut his finger, and I think it' unlikely that he could have licked it under the circumstances. Entry, we think, must have been through the cut. It is quite remarkable that, in this case, all'the various species should be eliminated from the body in faecal samples, because this presupposes that the thorium and the uranium must have been dissolved in body fluids and then eliminated from the body fluids into the gastro-intestinal tract. ■ With regard to the second question; we use a phantom rather than a point source for calibration and for stripping spectra. For all measure­ ments requiring high accuracy we use the spectra obtained under conditions of sim ilar sample volume and geom etry. INTERPRETATION OF DATA (Session 7)

INTERPRETATION OF BIOASSAY DATA

. G. W. DOLPHIN AND S. JACKSON HEALTH PHYSICS AND MEDICAL DIVISION, ATOMIC ENERGY RESEARCH ESTABLISHMENT, HARWELL, BERKSHIRE, ENGLAND

Abstract — Résumé — Аннотация — Resumen

INTERPRETATION OF BIOASSAY DATA. In order to estimate the radiation dose to the critical organ due to internal contamination with a radionuclide, it is necessary to measure the amount of radionuclide in the body. When the radionuclide emits gamma- or X-radiation, or beta-radiation giving rise to suitable bremsstrahlung, direct measurements of the body content can be made in a whole-body activity measuring facility, but when there is no suitable emission of radiation it is necessary to resort to estimation of the body, content from measurements of the amount of radionuclide in excreta, blood, breath or swabs taken from the nose or mouth. There has been some study of the excretion of radionuclides in urine as an index of internal contami­ nation, and urine analysis is the most extensively practised form of bioassay. The small amount of reference data available about the metabolism of radionuclides is still, however, a very serious weakness and a great deal more research is urgently necessary. In this review three broad categories have been used in classifying the different modes of metabolism of various radionuclides. The first group includes those radionuclides which are rather uniformly distributed throughout the body. The second group includes those which are concentrated particularly in one or more organs of the body. The third group, really a specially important sub-group of the second, comprises the bone seekers. The pattern of urinary excretion and particularly its relationship to the radiation dose delivered to the critical tissues, is distinctly different for these different groups. It is emphasized that, following the recognition of an accidental intake, it is desirable to analyse a carefully planned series'of urine samples, to provide a measure of the urinary excretion rate for a suitable period. The uncertainty associated with each individual result because of fluctuation in excretion rate can thus be minimized. A somewhat different approach is suggested in the case of urine sampling for routine surveillance of a group of personnel. The data available about metabolism of the radionuclide are used to evaluate an in­ vestigation level, namely, the urinary excretion rate corresponding to a chosen level of body content. It is important that measurements for routine surveillance should be made with maximum accuracy and carer fully recorded. These measurements may be used much later in making an assessment of .the chronically- acquired body content. A summary is presented of the human data available on metabolism and urinary excretion of tritium, caesium, uranium, strontium, radium and plutonium.

INTERPRETATION DES DONNÉES DE L’ANALYSE D’ÉCHANTILLONS BIOLOGIQUES. Afin d’évaluer la dose de rayonnements à l’organe critique due à la contamination interne par un radionucléide, il faut mesurer la quantité du radionucléide dans le corps. Quand le radionucléide émet des rayons gamma ou X, ou des rayonnements bêta donnant.naissance à un rayonnement de freinage convenable, on peut procéder à-desmesures directes de la charge corporelle de ce radionucléide dans un anthroporadiamètre, mais lorsque le radionucléide n’émet pas de rayonnements convenables, on est obligé d'évaluer la charge corporelle d'après les mesures de la quantité du radionucléide contenue dans les excreta, le sang, l’haleine ou des tampons provenant du nez ou de la bouche. Certaines études ont été faites sur l'élimination urinaire des radionucléides, considérée comme un indice de la contamination interne, et l’analyse de l'urine est la forme la plus répandue de l’analyse d’échantillons biologiques. Cependant, le caractère fragmentaire des données de référence dont on dispose sur le métabolisme des radionucléides constitue toujours une faiblesse très grave et il faut faire d’urgence beaucoup de travaux de recherche dans ce domaine. " Dans cette étude, on a classé les différents modes de métabolisme des divers radionucléides en trois grandes catégories. Le premier groupe comprend les radionucléides répartis dans l’ensemble du corps de façon

329 3 30 G. W. DOLPHIN and S.JACKSON

assez uniforme. Le deuxième groupe comprend ceux qui se concentrent particulièrement dans un ou plusieurs organes. Le troisième groupe, qui est en fait un sous-groupe particulièrement important du deuxième, com­ prend les ostéophiles. Le régime de rélimination urinaire, et notamment sa relation avec la dose de rayonne­ ments reçue par les tissus critiques, sont nettement différents pour ces divers groupes. Les auteurs insistent sur le fait qu'après la constatation d'une absorption accidentelle il convient d’analyser une série soigneusement constituée d’échantillons d’urines, afin d’obtenir une mesure du taux d’élimination urinaire pour une période convenable. On peut ainsi réduire au minimum le degré d’approximation inhérent à chaque résultat individuel, dû aux fluctuations du taux d’élimination. Dans le cas de prélèvements d’échantillons en vue de la surveillance régulière d'un groupe d'employés, on prppose d'aborder la question de façon quelque peu différente. ' On utilise les données dont on dispose sur le métabolisme du radionucléide pour évaluer un niveau d’alarme, à savoir le taux d’élimination urinaire correspondant à un niveau choisi de la charge corporelle. Il importe d'effectuer avec le maximum de précision les* mesures destinées à la surveillance régulière et de les noter soigneusement. Ces mesures peuvent être utilisées beaucoup plus tard pour faire une évaluation de la charge corporelle résultant d’une exposition chronique. . Le mémoire contient un résumé des données dont on dispose sur le métabolisme et rélimination urinaire du tritiu m , du césiu m , de l’uranium , du strontium , du radium e t du plutonium chez l'h om m e.

. ИНТЕРПРЕТАЦИЯ ДАННЫХ БИОЛОГИЧЕСКИХ ИССЛЕДОВАНИЙ. Для оценки дозы облучения критического органа, связанной с внутренним заражением радиоизотопами, необ­ ходимо определять количества изотопов в организме. Если радиоизотоп является источни­ ком гамма- или рентгеновых лучей или бета-излучения, вызывающего соответствующее тор­ мозное излучение, прямые измерения содержания его б организме могут производиться с помощью установки для измерения радиоактивности всего организма. Когда же соответствую­ щие излучения отсутствуют, определения-производятся с помощью измерения количества ра­ диоизотопов в выделениях, крови, вдыхаемом воздухе или мазках, взятых из носовой или ротовой полости. . . Проводилось изучение выделения радиоизотопов в моче как показателя внутреннего заражения. Анализ мочи является наиболее широко используемой формой биологического анализа. .Очень верьезным недостатком все еше является малое количество данных об обмене радиоизотопов, и необходимы дальнейшие исследования. . В данном обзоре для классификации различных видов обмена разнообразных изотопов использованы три большие группьк Первая группа включает*те радиоизотопы, которые доволь­ но однородно распределяются в организме. Вторая группа содержит те, которые концентри­ руются главным образом в одном или более органах тела. Третья группа фактически являет­ ся особо важной подгруппой второй группы и содержит откладывающиеся в костях изотопы. Выделение с мочой и особенно его связь с дозой, получаемой критическими тканями, отчет­ ливо различается для изотопов каждой из этих групп. Подчеркивается, что после диагностики случайного попадания в организм изотопа жела­ тельно провести серию анализов мочи, чтобы измерить скорость выделения изотопа с мочой за соответствующий период времени. Таким образом, неточности в каждом отдельном случае, связанные с колебанием скорости выделения, могут быть сведены к минимуму. Несколько другим может быть подход в случае взятия проб при обычном обследовании группы персонала. Имеющиеся данные относительно обмена радиоизотопов используются для оценки пограничного уровня, а именно — скорости выделения с мочой, соответствующей из­ бранному уровню содержания в организме. Важно, чтобы измерения при обычном обследова­ нии производились с максимальной точностью и тщатёльно регистрировались. Эти измерения могут быть использованы в значительной мере для оценки хронического поступления изотопа. Представлены выводы по имеющимся данным об обмене и выделении с мочой трития, цезия,,урана, стронция, радия и плутония у человека.

INTERPRETACIÓN DE DATOS OBTENIDOS POR BIOANALISIS. Para determinar la dosis de irradiación en el órgano critico, debida a contaminación interna por un radionúclido, es preciso medir la cantidad del mismo contenida en el organismo. Cuando dicho radionúclido emite rayos gamma, X o radiaciones beta que generan radiaciones de frenado, es posible medir directamente la carga corporal con un antropogammá- metro, pero cuando las radiaciones emitidas no se prestan para ello, hay que calcular la carga corporal basán­ dose en la medición de la cantidad de radionúclidos contenida en las excreciones, la sangre, el aliento o en frotis tomados de la.nariz o de la boca. . INTERPRETATION OF BÎOASSAY DATA 331

Se ha estudiado con bastante detalle la excreción de radionúclidos por vfa urinaria como indice de la contaminación interna, siendo el análisis de orina el procedimiento de bioanálisis más comúnmente practicado. Con todo, la escasez de datos disponibles sobre el metabolismo de los radionúclidos constituye aún un grave inconveniente y sé estima necesario realizar con carácter urgente nuevas y amplias investigaciones. En la memoria, las diferentes modalidades de metabolismo de los radionúclidos se clasifican en 3 grandes grupos. El primer grupo comprende los radionúclidos que se distribuyen de manera relativamente uniforme por todo el organismo. El segundo incluye los que se concentran particularmente en uno o más órganos. El tercer grupo, que constituye en realidad un subgrupo particularmente importante del segundo, comprende los radionúclidos osteófilos. El régimen de excreción por vfa urinaria y, en particular, su relación con la dosis de irradiación recibida por los tejidos críticos es netamente diferente para cada uno de estos grupos. Se recalca que una vez detectada una absorción accidental, conviene analizar una serie de muestras de orina tomadas con arreglo a un plan detallado, para poder medir la velocidad de excreción por vía urinaria durante un período adecuado. De este modo se consigue reducir al mínimo la indeterminación inherente a cada resultado aislado, debida a fluctuaciones de la velocidad de excreción. En el caso del muestreó de la orina para la vigilancia regular de un grupo de personas; se sugiere un método algo diferente. Los datos disponibles acerca del metabolismo del radionúclidó se utilizan para calcular un nivel de investigación, a saber, la velocidad de excreción por vía urinaria correspondiente a un nivel determinado de la carga corporal. Es esencial que las mediciones para la vigilancia regular se efectúen con la máxima precisión y se registren minuciosamente. Tales mediciones pueden utilizarse mucho tiempo después para evaluar la carga corporal crónica. . Se resumen los datos disponibles acerca del metabolismo humano y de la excreción por vfa urinaria del tritio, cesio, uranio, estroncio, radio y plutonio. .

1.- INTRODUCTION, ■

- The possibility of internal contamination of personnel working with radio­ nuclides is'minimizeid by enclosing processes and by maintaining environ­ mental contamination at a low level; surveys of surface and airborne con­ tamination are made to confirm that the necessary standards are being met. Because of the difficulties'of relating environmental conditions to the intake of individuals it is desirable to confirm the adequacy of these precautions by making a more direct measurement of the actual body content of the personnel concerned in order to estimate the radiation dose to the critical tissues. The best method of measuring the body content for those radionuclides which emit X- or gamma-radiation, e.g. caesium-137, radium-226, or bremsstrahlung, e.g. phosphorus-32, strontium-90, is by means of a whole- body-activity measuring facility, IAEA Symposium on Whole-Body Counting [1] and MEHL and RUNDO [2] . F o r the estimation of body content of these radionuclides which do not emit suitable penetrating radiations it is necessary to depend on bioassay, i.e., analysis of spécimens of tissue or excreta for their radionuclide content. Examples of radionuclides in this category are plutonium-239, natural uranium, and tritium . • In addition to the regular bioassay measurements made on personnel who are routinely working with radionuclides, special series of bioassay measurements are made on individuals who are known to have suffered an acute accidental intake. The primary purposes of these two types of bio­ assay investigation are different, as will be brought out in the course of this paper. Some of the problems of bioassay have-been reviewed by LANGHAM [3], WILLIAMS [4], LISTER [5], and JACKSON and TAYLOR [6]. 332 G. W. DOLPHIN and S. JACKSON

2. AVAILABILITY AND SUITABILITY OF BIOLOGICAL SPECIMENS FOR BIOASSAY

2.1. Nasal swabs

Only very low levels of internal contamination have usually been found in numerous personnel examined in a whole-body-activity measuring facility after detection of gamma activity on nasal swabs, HESP and SCHOFIELD [7] . Although results from nasal swabs do not necessarily give a good indi­ cation of the body content of radionuclides it is obvious that a frequent incidence of Heavily contaminated nose swabs reveals an unsatisfactory situation. EAKINS and MORGAN [ 8 ] have found better correlation between the amount of contamination in nose blows and the amount of radionuclide excreted in faeces in the three or four days following a known inhalation.

2. 2. Saliva and sweat

Shortly after intake by ingestion or inhalation, the amount of a radionuclide in the saliva may be a useful indication of the magnitude of the intake, but the amount retained in the body will depend on the degree of absorption from the GI tract and the lungs. After the inhaled or ingested radionuclide has been cleared from the upper respiratory tract and the gut, the saliva may conceivably be suitable material for the estimation of body content by bio­ assay. However, very little is known about the relationship between body content and saliva content of elements which do not play a part in normal physiology, and the collection and analysis of saliva present obvious problems. This also applies in the case of sweat, and the relationship of sweat content to body content is, fo r many elem ents, unknown. A special case, however, is tritiated water which rapidly diffuses throughout the water of the body. The concentration of tritiated water is therefore the same in any specimen of body fluid, including saliva or sweat. .

2.3. Breath

Breath analysis is restricted in application to the noble gases, notably radon and thoron, because volatile compounds of radionuclides are not commonly produced under physiological conditions. Radon in breath is used routinely as a measure of the radium content of the body, GROVE and CLACK [9]. HURSH and LOVAAS [10] have recently described a device for measuring thoron in breath and its use in determining body contents of thorium. It seems possible that C14O2 in breath may be useful for estimating carbono 14 in the body, DANCER et al. [11] . In principle, breath analysis could be used to measure the amount of exhaled tritiated water but it is more con­ venient to make measurements on urine samples.

2 .4 . B lood

Blood is the only tissue which can be sampled at all readily, and even in this case, there is a fairly severe limitation on the frequency with which samples can reasonably be requested from an individual. Most radionuclides INTERPRETATION OF'BIOASSAY DATA 333 are not retained predominantly in the blood, and the ratio of blood concen­ tration to the concentration in the critical organ is continuously changing. An exceptional case is that of phosphorus-32, after intravenous injection as the compound DFP (diisopropylfluorophosphonate); because most of this compound is bound by the blood cells GARBY [12] a single blood sample gives a useful measure of the total body content. Blood samples are less susceptible to accidental contamination than are most other bioassay samples.

2. 5. Faeces

The collection of faecal samples presents less practical difficulty than in the case of the other specimens so far discussed, but it is aesthetically objectionable to some people. Radionuclides may reach the faeces in several different ways. Apart from processes of, diffusion or active transport through the gastrointestinal wall, elements may be secreted with the digestive juices or the bile, the latter probably being the primary route of excretion for material deposited in the liver; also, much of the inhaled material initially deposited in the respiratory system is transferred to the GI tract. The existence of these four very different pathways makes it very difficult to interpret measure­ ments of radionuclides in faeces in terms of body content. Measurement of the amount of a radionuclide excreted in the faeces is useful in the case of elements which are not absorbed from the GI tract. Following ingestion it may be used to estimate the radiation dose to the c r iti­ cal portion of the gastro-intestinal tract. Following a single inhalation of insoluble material it may be used to estimate the amount of radionuclide retained deep in the lung (beyond the ciliated epithelium). In the ICRP (1959) [13] model for inhaled insoluble particulates, the amount transferred to the, GI tract (62|%) is five times that retained in the lung (12i%). By application of this model the faecal measurements can be related to the amount retained in the lung. The total amount of radionuclide excreted in the faeces during the first three or four days after the inhalation probably represents all that which was initially deposited on the ciliated epithelium ofthe tracheo-bronchial tree, the throat, and the nasal passages. However, tt should be emphasized that the size distribution of the inhaled particulate w ill determine the fraction of radionuclide which is deposited beyond the ciliated epithelium so that the relationship between the amount retained in the lung and the amount excreted in the faeces probably varies widely depending on the specific circumstances of the inhalation.

2. 6. Urine ;

The ready availability of urine inevitably makes it the most important material for bioassay, in spite of the considerable difficulties of interpre­ tation which are to be discussed in Sections 4 and 5. In a normal man about 170 1 of fluid per day is filtered from the plasma through the glomeruli of the kidney. Almost all the water is resorbed into the body via the kidney tubules leaving only about 1.4 1 of urine per day. In 334 G. W. DOLPHIN and S. JACKSON

the tubules selective resorption of soluble substances which have passed through the glomeruli occurs by active transport. There is also some active secretion from the plasma through the kidney tubules into the urine. Hence the exact relationship between' the concentration of soluble substances in the urine and in the plasma depends on the amount of active resorption and se­ cretion in the tubules.

3. ' GENERAL CONSIDERATIONS ABOUT THE METABOLISM OF RADIONUCLIDES AND THE RADIATION DOSIMETRY FOR INTERNAL CONTAMINATION

The distribution and metabolism of a radionuclide entering the body depends in the first instance on whether it is in the form of a compound which is soluble in body fluids. Radionuclides in the form of compounds which are insoluble in body fluids may be retained for a long time at the site of entry, such as the lungs or a wound; in such circumstances it is difficult to gener­ alize about the relationship between the urinary excretion rate and the amount of radionuclide in the body, a problem which w ill be discussed in Section 3.2. Attention will first be given to the case of those radionuclides which are in a form soluble in the body fluids. '

3. 1. Radionuclides soluble in body fluids

A simplified compartment model to represent the metabolism of a radio­ nuclide soluble in body fluids is given in Fig. 1. Radionuclides may enter the plasma or interstitial fluids by absorption from the GI tract or the lungs, or thoough a wound. Some of the radionuclide may then be preferentially concentrated in one or more organs, such as the liver, thyroid or bones, some may be excreted in the urine and faeces (also, to a smaller extent, in the sweat, and, in some cases, in the exhaled air). The object of making measurements of the amount of radionuclide in the excreta is to determine the radiation dose to the body tissues due to the ámount of the radionuclide which remains inside the body. . In general terms and for the purposes of this review these radionuclides may be divided into three groups according to their distribution and metabo­ lism in the body; A. Radionuclides which are not concentrated in any organ or in the red cells, e.g. sodium and bromide ions, which are diffused throughout the plasma and interstitial fluid, and tritium, in the form of tritiated water, which is 1 uniformly distributed throughout all the body fluids, intracellular as well as extracellular. B. Radionuclides which are deposited predominantly in one (or more) of the organs4of the body, è. g. iodine, which is concentrated in the thyroid gland, or which are concentrated in intra-cellular fluids, e.g. potassium and caesium. C. Bone-seeking radionuclides. Because of the major importance of irradi­ ation of the skeleton, radionuclides which are deposited predominantly in bone will be considered as a separate group. ' The characteristics of the above groups are discussed below in more detail. INTERPRETATION, ÔF BIOASSAY DATA 335

FAECES

■ Fig. l . ' ■ ■ ■ ’ .

■ A simplified compartment model illustrating metabolic pathways for ' inhaled or ingested radionuclides

Group A '

Radionuclides of this group are excreted exponentially. A good estimate of the body content can be made from a single sample of extra­ cellular fluid, or, more conveniently, of urine. In'the particular case of tritiated water, measurements of its concentration in any physiological fluids, e.g! urine, sweat or saliva, can be used to estimate the,concentration of tritium in the body water. In order to calculate the dose from such a single measurement it is only necessary to know the effective h alf-life of the radio­ nuclide in the body, the mass of tissue irradiated and the time at which the activity entered the body. .

Group В . .

Radionuclides in this group initially behave like those in group A, and diffuse.into the plasma and interstitial fluids. From these fluids the radio­ nuclide may follow a metabolic pathway leading to concentration in one or more of the body organs, or it may be excreted. The excretion pattern of this group of radionuclides is illustrated by the curve in Fig. 2 in which the daily excretion rate is plotted against time' elapsed since intake. . The initial rapidly-falling excretion rate, ab, corresponding to the clearance of the radionuclide from the body fluids, is followed by a slowly falling excretion rate, .be, corresponding to the slower release of the radionuclide from the 336 G. W. DOLPHIN and S. JACKSON

F ig . 2

Illustration, of a urinary excretion pattern of a radionuclide in group В

body organs. The part of the curve labelled ab cannot readily be related to the amount of radionuclide in the body organs where it is being concentrated and is of no real value in estimating the radiation dose to these organs. The part of the curve, be, is related to the release of the radionuclide from the body organs where it was concentrated and consequently if this part of the curve can be established from measurements of excreta it may be possible to make an estimate of the radiation dose to the organ.

Group С

This group comprises the bone-seeking radionuclides which present different problems from those discussed above in group B. The pattern of excretion is roughly the same as that shown in Fig. 2 but the second phase of the excretion falls off much more slowly due to the very long term re­ tention of any radionuclide which is incorporated in the bone. F or long-lived radionuclides this long-term retention in bone is best represented by a power function, DOLPHIN and EVE [14], which might be expected from the fact that the skeleton consists of many different types of bone where the bone mineral is turned over at different rates, BRYANT and LOUTIT [15]. The urinary excretion rate is consequently also best represented by a power function. Although there is a constant relationship between the urinary ex­ cretion rate and the body content when the excretion rate is represented by an exponential function, this is not the case when a power function applies. Another important difference between this group and group В is that the urinary excretion rate of radionuclides incorporated in the bone fluctuates more widely from day to day. Such fluctuations have been observed by MULLER et al. [16] in cases of strontium-90 contamination where the vari­ ation was as much as a factor of four during the period of about a week. HARRISON [17] also reports similar fluctuations in the excretion following a single intake of radiostrontium. These fluctuations are probably a con­ sequence of the lack of uniformity in rate and location of the accretion and INTERPRETATION OF BIOASSAY DATA 337

resorption processes continually taking place in the skeleton, but it has been suggested, HARRISON [17], that daily variation in the dietary intake of some elements, particularly phosphorus, also affects the excretion of strontium. Daily variation in the excretion of lead, another bone seeking element, has long been known to occur and was recently discussed in a review by KEHOE [18] . These fluctuations constitute a problem in the interpretation of urinary excretion data because a single measurement of excretion rate based on a small sample of urine may differ considerably from the mean excretion rate. This problem is further discussed in Section 4.

3. 2. Radionuclides insoluble in body fluids

In dealing with internal contamination of the-body with radionuclides, it is common practice to consider different compounds under the two cate­ gories "soluble" and "insoluble". It is important to note that the relevant consideration is the degree of solubility of a compound in body fluids which may be different from its solubility in non-physiological fluids. An im ­ portant example of this is plutonium nitrate, which, although freely soluble in acid, is hydrolysed to the insoluble hydrated oxide at physiological (neutral) pH. On the other hand, strontium carbonate, which is almost completely insoluble in water, has been shown by RUNDO and WILLIAMS [19] to be rapidly absorbed from the lung into the body fluids. ' After intakes of insoluble compounds, measurable urinary excretion of radionuclides commonly occurs due to the slow passage of material into the body fluids either in true or colloidal solution. The radionuclides in true solution should conform to the pattern of metabolism and excretion found for soluble compounds but any large colloidal particles will be deposited inthelymphnodes, spleen, liver, and bone marrow as a result of phagocytosis. It is obviously very difficult to estimate the content of radionuclide of any specific tissue from the urinary excretion rate. However, if all the intra­ cellular colloidal m aterial is slowly dissolving in the body fluids at the same rate, then the urinary excretion rate may be related to the total body content. There is a small amount of published data available about the urinary excretion resulting from the inhalation of insoluble compounds of radio­ nuclides, e.g. radium sulphate, MARINELLI et al. [20], uranium oxide, FISH [21] and SAXBY et al. [22], strontium carbonate, RUNDO and WILLIAMS [19], plutonium oxide, DOLPHIN [23] and Eakins and Morgan [8]. Among these cases there is a wide variation in the relationship between the urinary excretion rate and the body content. Further progress in this problem on estimating body content from urinary excretion rate must depend on seizing every opportunity to collect more data from human cases of intake and from animal experiments. ’

4. INTERPRETATION OF URINE ANALYSIS MEASUREMENTS FOR INDIVIDUAL CASES FOLLOWING A SINGLE INTAKE OF RADIONUCLIDE

When an acute accidental intake of radionuclide is known to have oc­ curred, the prim a ry object of any subsequent investigation by environmental measurement, in vivo measurement and bioassay is to estimate the radiation

22 . 3 38 G. W. DOLPHIN and S. JACKSON dose to the tissues of the body and, in particular, the critical organ. In this investigation a well-planned series of measurements of the urinary excretion rate of the radionuclide must play a very important part. In order to determine the urinary excretion rate it is necessary to measure two quantities: (a) the amount of radionuclide in the sample of urine; and (b). the period of metabolic activity represented by the sample of urine. The radiation dose can then be estimated from the measurements of urinary excretion rate, provided there is information on the following: (c) the relationship of urinary excretion rate to body content; and (d) the distribution of the radionuclide in the body and thé fraction present in the critical organ. These measured quantities and relationships will be considered in more detail. • ' ■ In general, good analytical techniques have been developed and the determination of the amount of radionuclide in a sample of urine can be made reliably, Jackson and Taylor [6]'. This is the strongest link in the chain. However^ the reliability of the analytical results depends on the successful exclusion of accidental contamination of the urine samples during and after voiding and strict control of the traces of radioactivity introduced by the chemical procedures. It is surprisingly difficult to define the period of metabolic activity re­ presented by a sample of urine. One method is to collect all the urine ex­ creted in a 24-h period, in which case the bladder must be empty at the beginning and end of the period. Unless the contaminated person is unusually co-operative and well-disciplined this method can, in practice, by unreliable because it involves collection of samples when outside the laboratory or factory. To avoid the necessity of collecting samples away from the place of work, other methods based on sample volume or creatinine content may be used. The volume of a urine sample is not a very reliable index because it depends on environmental and physiological conditions, such as fluid intake, ambient temperature and physical activity. The ICRP Committee II (1959) [13] takes the daily urinary excretion of standard man as 1.4 1. Probably the most reliable index is the creatinine content in the urine sample. JONES [24] has shown that there is a much sm aller degree of variability in creatinine excretion rate than in sample volume. The mean rate of creatinine excretion is about 1.7 g/d in men and 1.0 g/d in women. Jones [24], MANSERGH and RILEY [25], MULDOWNEY et_al. [26], BAZZANO and GHISLANDI [27]. . The relationship between urinary excretion rate and body content is based for most radionuclides on a small number of case histories of human metabo­ lism. There is considerable variation between individuals, and more data on human metabolism of the more important radionuclides are urgently needed. The degree of uncertainty about the significance of each individual sample analysis is reduced if a series of results is available. ' When an estimate of the body content of a radionuclide has been made it is then necessary to calculate the radiation dose delivered to the critical organ. This calculation depends on a knowledge of the distribution of the' radionuclide in the body and in particular the fraction which is located in the critical organ. One source of such human data is post-mortem exami­ nations carried out following the accidental death of radiation workers. These INTERPRETATION OF BIOASSAY DATA 339

data are of limited value because the levels of contamination are usually very low and it is difficult to trace the times when intakes occurred. Some data have been obtained in the United States from planned experiments on terminal hospital patients. The situation is entirely different if the radio­ nuclide emits suitable gámma-or X-radiation, e.g. caesium -137, strontium-85, uranium-235 and iodine-131, when the amount present in a living person can be directly measured and possibly located in a body organ. ' In the foregoing discussion the object has been to determine the radiation dose to the critical organ. This is practicable and desirable in the case of radionuclides with short effective half-lives in the body. In the case of radionuclides with long effective half-lives, the total dose is to a close ap­ proximation, proportional to the dose-rate, and it is usual to estimate the body content on which the dose-rate depends. The estimated body content is compared with the maximum permissible body burden, ICRP (1959) [13], which delivers the maximum permissible dose-rate to the critical organ.

5. URINE ANALYSIS IN THE ROUTINE SURVEILLANCE OF GROUPS OF RADIATION WORKERS

A distinction must be drawn between the routine surveillance of a group of workers exposed to radionuclides which are short-lived in the body, e. g. tritium, and those which are long-lived in the body, e.g. plutonium. In the case of those nuclides which are short-lived in the body, the available data about urinary excretion are used to evaluate an investigation level. The investigation level is set so that if results of urine analysis fall below this level no action need be taken. Usually this level is chosen to correspond to a body content of 1/lOth of the maximum permissible body burden (mpbb) recommended by the ICRP (1959) [13] . Because most radio­ nuclides with a short effective life in the body are excreted exponentially, the same investigation level may be used for a given radionuclide throughout the period of work, no matter how long. If the level is exceeded consistently over a period of time or seriously exceeded on any one occasion, then a decision must be taken about the necessity for further investigation or other special action. In making this decision consideration must be given to the difference in the radiation doses delivered to body tissues by different radio­ nuclides. For example, an initial retention of 1 mpbb of tritiated water delivers 0.2 rem to the body water, 90% of this dose in 36 d, whereas an initial retention of 1 mpbb of caesium-137 delivers a dose of 1.35 rem to the total body, 90% in 210 d. (These doses are calculated using the concepts and data from ICRP (1959) [13]). When it has been confirmed by means of one or more measurements on urine samples that the investigation level has been seriously exceeded, a special sampling programme is initiated in order to assess the radiation dose, as described in Section 4. An investigation of the environmental conditions in the working area should also be made in order to ascertain the cause of the intake. . In the case of bone-seeking radionuclides (group С of Section 3.1) with a,, long effective life in the body, the long-term urinary excretion rate does not follow an exponential law but is more satisfactorily represented by a power law. In consequence, there is not a constant relationship between the 3 40 G. W. DOLPHIN and S. JACKSON urinary excretion rate and the body content, even following a single intake. This relationship becomes still more difficult to define if more than one intake has occurred. Hence it is not sufficient to compare each urine ana­ lysis result with the investigation level; all the previous results must also be brought into consideration in order to make a provisional estimate of the body content. The investigation level for these radionuclides is of necessity chosen somewhat arbitrarily. The frequency with which routine samples are taken depends primarily on the degree of confidence in the containment or other means of control of environmental contamination. As confidence increases with successful oper­ ation of a particular installation or working procedure the frequency of routine samples may be reduced. Even if there is a high degree of confi­ dence in the containment,. urine samples must be taken frequently enough to ensure the detection of any single intake large enough to deliver a signi­ ficant radiation dose. One important consideration in deciding the sampling frequency in this case is the effective life of the radionuclide in the body. A higher frequency of urine sampling is required for personneTengaged in special operations in which the environmental contamination inevitably cannot be controlled and measured as satisfactorily as in normal operations. From the above discussion it is obvious that the choice of sampling frequency in the particular factory or laboratory is a matter of judgment based on working experience. As already described in Section 3, there are considerable day-to-day fluctuations in the urinary excretion rate of bone-seeking radionuclides (group C). BEACH and DOLPHIN [28] give data which show the extent of these fluctuations for urinary excretion of plutonium. The most economical way of establishing the mean urinary excretion rate is to measure an aliquot of the combined urine samples collected over a period of several days rather than to make separate measurements on a number of daily samples. If urine samples are combined it is important to have an index of the total period of metabolic activity represented by the combined sample; the creatinine content (discussed in Section 4) appears to be the most practicable index for this purpose.

6. CONSIDERATION OF SOME IMPORTANT RADIONUCLIDES

In this section a more detailed review is made of data relevant to urine analysis for six important radionuclides, namely tritium, caesium, uranium, strontium, radium and plutonium.

6.1. Tritium

In the majo.rity of processes involving tritium, the tritium is' in the form of tritiated hydrogen gas or tritiated water. The solubility of hydrogen in body fluids is very small and the rate of oxidation to water within the body is very slow, PINSON and LANGHAM [29]; as a consequence the critical râdiological consideration fot~ tritium gas is irradiation of the skin. Urine analysis is not strictly relevant to the case of exposure to tritium gas but the possibility that tritiated water may be produced by oxidation can seldom if ever be excluded. INTERPRETATION OF BIOASSAY DATA 341

Tritiated water diffuses throughout the body water and because of the short range (l^rn mean range in water) of the low-energy beta particles, the critical,organ is the total body water, i.e.,the soft tissues. Tritiated water may enter the body by ingestion or by skin absorption; if the vapour is present in the atmosphere it is absorbed via the lungs and the skin, Pinson and Langham [2 9] . Provided that the fluid intake is not too rapid to permit equilibrium with all the body fluids before passage from the kidneys to the bladder, the concentration of tritiated water in the urine is equal to that in the body fluids. Hence, the estimation of body content from measurements of tritiated water in urine is very simple. In order to estimate the radiation dose delivered to the body as a result of an intake of tritiated water, it is necessary to know the rate of turnover of the body fluids. The biological half-life of tritiated water in the body is affected by fluid intake and varies between individuals. Values recently reported by WYLIE, BIGLER and GROVE [30] for six men ranged from 7.0 to 12.1 d, the average being 8.5 d. This is somewhat less than the values of. 9.5 d reported by RICHMOND et a l. [31] and 11.5 d reported by Pinson and,Langham [29] . ICRP (1959) [13] has adopted a conservative value of 12 d. The retention of tritiated water by the body may be written as follows

where R(t) is the percentage of the initial intake retained after t days. The effective half-life is 12 d, very similar to the biological half-life, because the physical half-life is very long (12.3 yr). This equation is related to the urinary excretion rate as follows

= a 5.8 exp (-0.058t), where U(t) is the percentage of the intake excreted in the urine after t days and a is the fraction of the total excretion which appears in the urine. Ac­ cording to the standard man values adopted by ICRP (1959) [13], a= 0.6, so that the equation may be written as

U(t) =3.5 exp (-0.0581).

This equation is shown graphically in Fig. 3. The urinary excretion rate expressed as a percentage of the body content at the time of excretion is constant at a value of 3.5% which is shown in Fig. 3. The maximum permissible body burden recommended by ICRP (1959) [13] is 1000/uc and the corresponding urinary excretion rate is Збдс/d. One- tenth of this, say 4|uc/d, is a satisfactory investigation level. In assessing the significance of a single urine measurement exceeding the investigation level it should be noted that even if the initial body content is as high as 5000yc only 1 rem will be delived to the critical organ. 342 G. W. DOLPHIN and S. JACKSON

(DAYS) .

' Fig- 3

Excretion of tritiated water following a single intake

If tritium is taken into the body in the form of organic compounds, the turnover rate may be slower than with tritiated water; even in the case of tritiated water a small fraction of tritium is retained for a much longer time, Pinson and Langham [29] . The mode of metabolism of each individual compound must be considered specifically. An important example is tritiated thymidine, which is incorporated in DNA; this leads to selective irradiation of cell nuclei. v

6.2. Caesium-137

Many studies of the distribution and metabolism of caesium have been reported in the literature and most of these have been reviewed in a paper on caesium-137 metabolism in man by ROSOFF et al. [32] . More recently studies on normal subjects have been reported by HESP, [33] RUNDO and T A Y L O R [34] and RUNDO [35] . The most commonly encountered compounds of caesium are soluble and are almost completely absorbed from the gut. Measurements on plasma following intravenous injection suggest that caesium diffuses into the extra­ cellular fluid within five minutes, Rosoff et al. [32]. Caesium in the extra­ cellular fluids is taken up by red cells and soft tissues and probably con­ centrated in the intra-cellular fluid like its physiological analogue potassium. Rough values for clearance rates from the extra-cellular fluid (approxi­ mately 20 1) can be calculated from the data given by Rosoff et al. [32] to be about 2 1/d via the kidney and 16 1/d to the red blood cells and soft tissues. These values are in agreement with an analysis made by Rundo [35] of the available data on retention of caesium in which he found a mean value of about 10% (ranging from 6 to 15%) for the amount of caesium retained with INTERPRETATION OF BIOASSAY DATA 343 a half-life of 1-2 d; the remainder was retained with a mean half-life of 110 d (ranging from 50 to 150 d). From the above discussion it is possible to write down an equation in round numbers for the retention of caesium -137 in the body as follows

„ Í 0.693 V f 0.693 , R(t) = 10 exp ( ------j— t) + 90 exp ( — ^ q t where R(t) is the percentage of the intake retained at t days after intake. The effective half-lives have the same value as the biological half-lives because the physical half-life for caesium is comparatively long (30 yr). This equation may be related to the urinary excretion rate as follows

. . dR U(t)=a ж

0.693 = a 6.9 exp + a 0.62 exp t

where U(t) is the percentage of the intake excreted in the urine on day t and a is the fraction of the total excretion which appears in the urine. Rosoff et al. [32] found that the ratio of urinary to faecal excretion was 10:1 and Rundo and Taylor [34] found ratios ranging between 9 : 1 and 4 : 1 with a median about 7:1. In this review a is taken to be 0.9. The equation is graphed in Fig. 4 and shows that between 10 and 100 d after intake the mean urinary excretion rate ranges from about 0.5% to 0.3% of the intake per day. After about 10 d the urinary excretion rate expressed as a percentage of the body content at the time of excretion is constant at about 0.6%/d as shown by the horizontal line in Fig. 4. The maximum per­ missible body burden recommended by ICRP (1959) [13] is ЗОцс and hence the excretion rate corresponding to this body content is 0.18|ic/d. The in­ vestigation level can be taken as one-tenth of this excretion rate, namely, 0.018 д с /d or 0.02 цс in round numbers. In the first four or five days fol­ lowing an intake the urinary excretion rate will be high, as shown in Fig. 4, so that care must be taken not to use the investigation level as a criterion of urinary excretion rate for samples taken during the early phase of excretion. It should be noted that a simple whole-body-activity measuring apparatus might be the best way of routinely examining workers for internal contami­ nation with caesium-137.

6. 3. Uranium

Interpretation of the results of urine analysis for uranium has been re­ viewed by JACKSON [36, 37] . ■ BERNARD and STRUXNESS [38] measured the excretion of uranium by terminal brain cancer patients after intravenous injection of .soluble uranium compounds. Faecal excretion of uranium was 3 44 G. W. DOLPHIN and S. JACKSON

■ F i g -4 '

Excretion of caesium -137 following a single intake negligible, so that the daily urinary excretion as a percentage of the original body content is

U(t) = — = 6.9t'1-5. ' at

This equation is graphed in Fig. 5 together with the curve showing the urinary excretion rate as a percentage of the body content at the time of excretion. The relationships are very different because of the very rapid urinary ex­ cretion during the first day. With soluble compounds of natural uranium, the critical consideration is the non-radiological toxicity for the kidney. The maximum permissible concentration in air, (MPC)a, recommended by ICRP (1959) [13] is under­ stood to be based directly on data from animal experiments, VOEGTLIN and HODGE [39, 40] , so that it is appropriate to use the urinary excretion rate shortly after intake as a measure of the air concentration in which a man has worked and to cfalcula.te an investigation level in urine which corresponds to one-tenth (MPC)a. The information from the excretion curve which is used in the following computation is that two-thirds of the amount of uranium initially retained in the body is excreted in the urine during the next 24 h; the values for other quantities are taken from the Standard Man models of ICRP (1959) [13] . (M PC )a = 210 |Ug/m3 (soluble, natural uranium) Volume inhaled = 10 m3 (8-h shift) INTERPRETATION OF BIOASSAY DATA 345

. Fraction of inhaled particles retained and absorbed = 25% Fraction of initially retained uranium excreted during next day = I Volume of urine excreted per day = 1.4 1. Hence, the concentration of natural uranium in urine immediately after in­ halation of (MPC)a throughout a shift

25 2 1 = 210X 10 X y ^ X g -X у - j Mg/1 = 240 Mg/1­

In this case of non-radiological toxicity, extensive experience has shown that 1Ó0 Mg/1 of urine is a satisfactory investigation level for samples taken immediately after work with soluble natural uranium, BUTTERWORTH [41] .■

(DAYS) .

. F ig . 5 .

Excretion of soluble uranium following a single intake

Urine samples taken shortly after an intake of soluble uranium are most unsuitable for making an estimate of the body content because the urinary excretion rate is changing very rapidly at this time. In the case of highly- enriched uranium, with its greatly enhanced content of the isotope uranium- 2 34, increased irradiation of the bone is more critical than toxicity to the kidneys, and bone becomes the critical organ. For a considerable time after an intake of soluble uranium, any excretion due to release from the bone is overwhelmed by excretion of uranium released more rapidly from other tissues. Urine samples representing release of uranium from the bone can only be obtained after prolonged removal from the possibility of uranium intake. A sample taken after absence on holidy is probably the best available 346 G. W. DOLPHIN and S. JACKSON but it may be more practicable to take samples after a week-end, before resuming work. It should be emphasized that if the result for a sample taken at a shorter time after contact with uranium falls below the investigation level set for.after-holiday samples, there is a bigger margin of safety be­ cause the time of sampling has moved to a point when the excretion rate is higher. From consideration of the post-mortem data from their hospital patients Bernard and Struxness [38] concluded that the rate of turnover of uranium in bone could be defined in terms of a biological half-life of 300 d. This is in agreement with animal experiments, Voegtlin and Hodge [39, 40]. Hence about 0.2% o f the bone content is excreted (see Fig. 5) daily. The maximum permissible body burden recommended by ICRP (1959) [13] for uranium-234 is 0.05 juc, 85% of which is in the bone; the corresponding daily urinary excretion rate is 10'4/uc. One-tenth of this, namely 10"5^c/d, is suggested as the investigation level. The body content of uranium resulting from inhalation of insoluble urani­ um compounds is difficult to estimate from the results of urine analysis, Jackson [37] ; the published data on three human cases reveal a considerable variability in metabolism and excretion, FISH [21, 42] and SAXBY et a l. [22]. Further progress with this problem requires the collection of data from many m ore human cases.

6.4. Strontium-90

The metabolism of strontium in normal adults was reviewed by Dolphin and Eve [14] and they proposed a simple model to represent its metabolism. DOLPHIN and EVE [43] suggested that the equations representing the r e ­ tention of strontium in the body might be written in round numbers as follows

where R(t) is the percentage retained t days after the intake. The first term on the right-hand side of the equation shows that half the strontium is excreted rapidly from the body with a half-life of 2.4 d and the second term shows that 50% of the intake is retained according to a power law. The power law represents the long-term retention in bone. The fraction of the total excretion which is found in the urine is 0.8, so that the daily urinary excretion U(t) may be written as follows

This equation is shown graphically in Fig. 6 together with a curve which shows how the urinary excretion rate expressed as a fraction of the body content varies with time after the intake. The excretion rate falls off by a factor of about 20 during the first 20 d and during the next 100 d it falls off by a further INTERPRETATION OFÍBIOASSAY DATA 347

Fig. 6

Urinary excretion of strontium-90 following a single intake

factor of 10. This falling excretion rate makes it essential to know the day of intake if the initially retained amount is to be estimated following a single intake. A notable similarity of the strontium excretion curves in a number of different cases of strontium intake has been reported by Lister [5] . It is not possible to specify a single investigation level because of the non-exponential excretion and the data obtained from routine measurements must be periodically assessed as a whole in order to estimate the body content. The maximum permissible body burden recommended by ICRP (Г959) [13] is 2 juc. In ord er to ensure that less than one-tenth of the m axi­ mum permissible radiation dose has not been received during the preceding 13 weeks the investigation level is taken as lOOpc of strontium-90 per day in the urine. This investigation level is chosen for a three-month sampling interval but a higher level could be chosen for a more frequent sampling p rogram m e. One other important factor which must be taken into account when in­ terpreting measurements of strontium in urine is the large fluctuation in the amount excreted from day to day, Muller et al. [16] and Harrison [17] . As previously mentioned in Section 5, a representative sample can only be ob­ tained by collecting urine over a period of several days.

6. 5. R a d iu m -226 ■

V ery little information has been published about the urinary excretion of radium. There has been extensive study of the biological effects of radium deposited in man, in which estimation of the body content of radium has been based on measurement of radon-222 in the breath or on direct measurement 348 G. W. DOLPHIN and S. JACKSON of gamma-emission from radium-C (bismuth-214) in the body, EVANS [44] , Marinelli et al. [20], NORRIS et al. [45], RUNDO [46]. ’ The maximum permissible body burden of radium-226 is 0.1 цс, ICRP (1959) [13], by disintegration, this yields 1.26 X 10"5mc of radon-222 per minute and 70% of this is exhaled, Norris et al. [45] . The concentration of radon in the breath corresponding to 0.1 дс of radium-226 depends on the breathing rate of the individual but may be taken in round numbers as 1 pc/1 of breath. One-tenth of this concentration in the breath may be taken as the investigation level. From the early measurements of SCHLUNDT et al. [47] and from their own findings on 17 of Schlundt's patients who were still available for measure­ ment 21 years later, Norris et al. [45] obtained the equation

R(t) = 54 Г 0'52,

to describe the percentage retention of radium in the body for this long period. The daily total excretion rate is therefore

M î i =_28t-1-5 dt ’

Only about 2% of the excreted radium was in the urine, hence

U(t) = 0.56 t ' 1-5.

This equation is graphed in Fig. 7 together with the urinary excretion rate as a percentage of the body content at the time of excretion. - The rate of excretion of radium after a case of accidental inhalation of radium chloride was investigated by AUB et al. [48] during the period from 20 to 200 d after the intake and his findings are closely in agreement with the above equations. Because the excretion rate is not exponential, it is not possible to specify a single investigation level for radium in urine which is universally appli­ cable. For a three-monthly sampling programme, a level of 1 pc of radium-226 in urine per day ensures that less than one-tenth of the maximum permissible body burden is present. Six human cases of inhalation of radium sulphate have been reported by Marinelli et al. [20] but there are insufficient data about the urinary ex­ cretion rate to make it possible to draw any satisfactory conclusions about its relation to body content.

6.6. Plutonium -2 39

Plutonium does not have an analogue which is of physiological signifi­ cance. Although it might have been expected to show some analogy to urani­ um its metabolism, distribution and excretion from the body are strikingly different. DOLPHIN [49] has reviewed some of the problems connected with the interpretation of urinary excretion data for both single accidental intakes and chronic exposure. Plutonium is not absorbed from the gut to any significant INTERPRETATION OF BIOASSAY DATA 349

F ig - 7

Excretion of soluble radium following a single intake

extent; it can enter the body via the lungs or through a wound or burn. Except for specially-prepared solutions of plutonium citrate, all the commonly- encountered plutonium compounds are not properly soluble in body fluids and consequently tend to move slowly from the site of entry. This hold-up of the plutonium at the site of entry makes it very difficult to interpret .measure­ ments of plutonium excretion rate by reference to the data obtained by Langham [3] from controlled human experiments in which plutonium citrate .was injected intravenously. Beach and Dolphin [28] have calculated expected urinary excretion rates assuming that the plutonium is released slowly from the site of entry and the results of some of their calculations are shown in Fig. 8. Of the three curves in Fig. 8 one was obtained from Langham's [3] original data and represents excretion following a plutonium citrate injection. The other two curves were calculated, one on the assumption of a hold-up at the site of entry with an exponential release having a half period of 50 d and the other on the assumption that the release follows a power law with a first half period of 25 d. These curves show that the initial excretion rate following an intake is very dependent on the rate of release from the site of entry, After about a year the excretion rate appears to be less dependent on the rate of release and these curves indicate that a more reliable estimate of the body content may be possible at longer times after the intake. The investigation level is particularly difficult to choose for plutonium and although the curves given in Fig. 8 are very useful in indicating possible trends in urinary excretion rate it should be remembered that they are not directly founded on experimental data. A value of 0.4 pc of plutonium per day in the urine appears to be reasonable for the investigation level. 35 0 G. W. DOLPHIN and S. JACKSON

(DAYS)

F ig . 8

Utinaty excretion of plutonium following an intake of 0. 04 ¡sc

- The development of plutonium oxide as a reactor-fuel has drawn attention to the difficulty of estimating the lung content of inhaled plutonium oxide. Dolphin [23]- has suggested that the long-term urinary excretion rate may be related to the whole-body content in this case but more evidence is re­ quired in order to test the validity of this suggestion. At present, the best basis for the estimation of the lung content of insoluble plutonium following a known inhalation is the plutonium content of the faeces during the four days after the incident, see Section 2.5.

7. CONCLUSIONS

For soluble radionuclides there are some metabolic data available which may be used as a basis for the assessment of the internal contamination from, the results of urine analysis. In individual cases of accidental intake of soluble radionuclides, which can be investigated by a planned programme of urine analysis, it is possible to make an estimate of the radiation dose to body tissues. For routine surveillance of a group of radiation workers it is possible to derive an investigation level for the amount of radionuclide in the urine. The situation is much less satisfactory in the case of insoluble radionuclides for which it is only possible to make very tentative estimates of body contents and to set rather arbitrary values for investigation levels. More data are required about the metabolism of radionuclides, es­ pecially in the case of those which are insoluble in body fluids. Such data can be obtained from carefully-planned animal experiments and from ac- INTERPRETATION OF BÍOASSAY DATA 351 cidental cases of human intake when every effort should be made to collect and publish the maximum amount of information. Finally, the fact must be faced that the problems of estimation of the body content of radionuclides by bioassay will not be satisfactorily solved until more human volunteer experiments have been carried out.

ACKNOWLEDGEMENTS

The authors wish to thank Dr. I. S. Eve for his help in the preparation of this review and D r. K. P . Duncan for helpful discussions on the subject of urine analysis.

REFERENCES '

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Whole-Body Counting, IAEA, Vienna (1962). [2] MEHL, J. and RUNDO. J., Hlth Phys. 9 (1963) 607. [3] LANGHAM, W. H ., Brit. J. Radiol. Suppl. 7 (1957) 95. [4] WILLIAMS, K ., AERE R 3288 (1960). [5] LISTER, B. A. J. , "The problems and methods of sample assay", Diagnosis and Treatment of Radioactive Poisoning, IAEA, Vienna (1963) 23.

[ 6 ] JACKSON, S. and TAYLOR, N. A . , AHSB (RP) R37 (1964). ■ [7] HESP, R. and SCHOFIELD, G. B., Private communication (1964).

[ 8 ] EAKINS, J. D. and MORGAN, A. , these Proceedings (1964). [9 ] GROVE, W. P. and CLACK, B. N ., Brit. J. Radiol. Suppl. 7 (1957) 120. [10] HURSH, J. B. and LOVAAS, A. , Hlth Phys. 9 (1963) 622. [1 1 ] DANCER, G. H. , MORGAN, A. and HUTCHINSON, W. P. , AERE M 1190 (1963).

[12] GARBY, L ., Brit. J. Haematol. 8 (1962) 15. [13] INTERNATIONAL'COMMISSION ON RADIOLOGICAL PROTECTION, Publication 2 (1959) Pergamon Press, London (1960). .

[1 4 ] DOLPHIN, ,G. W. and EVE, I. S . , Phys. Med. Biol. 8 (1963) 193. [15] BRYANT, F. J. and LOUTIT, J. F. , AERE R 3718 (1961). [16] MULLER, J ., DAVID, A ., REJSKOVA, M. and BREZIKOVA, D ., Lancet П (1961) 129. [17] HARRISON, G. E. , Private communication (1964). [18] KEHOE, R. A. , Industrial Hygiene and Toxicology, Vol. II (PATTY, F. A. , Ed.), John Wiley & Sons, New York (1963) 941; [19] RUNDO, J. and WILLIAMS, K. , Brit. J. Radiol. 34 (1961) 734. [20] MARINELLI, L. D ., NORRIS, W. P ., GUSTAFSON,_ P. F. and SPECKMAN, T. W. , Radiology 61 (1953) 903. [21] FISH, B. , Inhaled Particles and Vapours (DAVIES, C. N. , Ed.), Pergamon Press, London (1961) 151. [22] SAXBY, W. N. , TAYLOR,. N. A. and GARLAND, J ., these Proceèdings (1964). [23] DOLPHIN, G. W. , these Proceedings (1964). [24] JONES, O. , WSL/417 (1952). [25] MANSERGH, J. and RILEY, C. J., ARDCA>267 (1957). [26] MULDOWNEY, F. P. , CROOKS, J. and BLUHM, M. M. , J. clin. Invest. 36(1957) 1375. [27] BAZZANO, E. and GHISLANDI, E. , Energía Nucleare 10 (1963) 507. [28] BEACH, S. A. and DOLPHIN, G. W. , these Proceedings (1964). [29] PINSON, E. A. and LANGHAM, W. H., J. applied Physiol. 10 (1957) 108. . [30] WYLIE, K. F ., BIGLER, W. A. and GROVE, G. R. , Hlth Phys. 9 (1963) 911. [31] RICHMOND, C. R. , LANGHAM, .W. H. and TRUZILLO, T. T ., J. cell. comp. Physiol. 59 (1962) 45. [32] ROSOFF, B., COHN, S. H. and SPENCER, H ., Rad. Res. 19 (1963) 643. [33] HESP, R. , these Proceedings (1964). [34] RUNIIO, J. and TAYLOR, B. T ., these Proceedings (1964). [35] RUNDO, J., Brit. J. Radiol. 37 (1964) 108. [3 6 ] JACKSON, S . , AHSB RP R15 (1 9 6 2 ). [37] JACKSON, S .,' these Proceedings (1964). 352 G. W. DOLPHIN and S. JACKSON

[38] BERNARD, S. R. and STRUXNESS, E.G ., ORNL 2304 (1957). [39] VOEGTLIN, C. and HODGE, H. C ., Pharmacology and Toxicology of Uranium Compounds, Pts. I and II, McGraw-Hill, New York (1949). [40] VOEGTLIN, C. and HODGE, H. C ., Pharmacology and Toxicology of Uranium Compounds, Pts. Ill and IV, McGraw-Hill, New York (1954). [41] BUTTERWORTH, A., Trans. Assoc. Indust. Med. Officers 5 (1955) 36. ' [42] FISH, В., HASL 58 (1958) 126.

[43] DOLPHIN, G. W. and EVE, I. S . , Phys. Med. Biol. 8 (1963) 205. [44] EVANS, R. D ., Amer. J. Roentgenol. 37 (1937) 368. [45] NORRIS, W. P. , SPECKMAN, T. W. and GUSTAFSON, P. F ., Amer. J. Roentgenol. 73 (1955) 785. [4§] RUNDO, J. , these Proceedings (1964). [47] SCHLUNDT, H. , NERANCY, J. T. and MORRIS, J. P. , Amer. J. Roentgenol. 30 (1933) 515. [4 8 ] AUB, J. .C ., EVANS, R. D. , GALLAGHER, D. M. and TIBBETTS, D. M. , Ann. internal M ed. 11 (1938) 1443, [49] DOLPHIN, G. W ., in press (1964).

. DISCUSSION

W.S. SNYDER: The general models you have presented are appropriate models for certain typical individuals and situations. I assume that you adjust the parameters to fit the data in an individual case, rather than re­ garding the model as fixed. G.W. DOLPHIN: The parameters are adjusted to fit the individual case when values are available from bioassay measurements. However, in assessing the results of routine bioassay measurements the best mean values available are fitted into the models. W.S. SNYDER: Some studies have suggested that the Pu excretion patterns occurring after large intakes may differ from those occurring after sm all intakes, in other words that they depend on the amount of Pu reaching blood. Do you have any intake levels to suggest as appropriate to the various curves? , G.W. DOLPHIN: Unfortunately I have no information on this matter. R.T. MOORE: Do the models, as shown, refer to normal physiological function alone, or do they also refer to kidneys with a pathological condition? G.W. DOLPHIN: These models are taken from the standard text-books on physiology. J. MUELLER: I would like to say something about faecal analysis. It is rather difficult to deduce lung burdens due to the inhalation of insoluble radioactive aerosols, because there are always two fractions: a fraction which has been swallowed and a fraction which has been inhaled. I think we very rarely have pure inhalation accidents, or pure ingestion accidents. Just a few days ago I dealt with six people involved in an accident with a 1 promethium dye. Although there was very little difference in the urine levels, the faeces-to-urine excretion ratio in those six cases differed by three orders of magnitude during the first few days. The different cases must therefore have ingested different fractions of the material. We also found that on the first day one of the cases excreted about three orders of magnitude m ore than the others, but on the tenth day their excretion in faeces and urine was similar. That can only mean that the first man ingested a rather large amount, which passed through his gut without being absorbed. I think, there- INTERPRETATION OF BIOASSAY DATA 353

Correlation between urinary Sr90 and Ca six years after chronic Sr90 contamination fore, that simply making a faecal analysis for the first few days and estimating the lung burden on that basis is always rather a difficult problem. The second point I would like to make concerns the variation in excretion of strontium from day to day. Figure ID shows some data we recently re­ ceived of urinary excretion of strontium and calcium in cases of chronic strontium contamination dating six years back. We find that the day-to-day variation is considerable, but the strontium and calcium excretion vary together. These cases have been under observation now for six years, and we have calculated body burdens on the basis of long-term excretion analysis.

F ig . 2D

Urinary Sr^/Ca ratio as a function of estimated Sr90 whole-body burden 354 G. W. DOLPHIN ' and S; JACKSON

Figure 2D shows the Sr/Ca ratio versus body burden as estimated from the long-term excretion studies, and again we get a very good correlation. As a matter of fact, the correlation coefficient was 0.99, which is really too good to be true, but that is what our statistician tells me. In these cases of long-standing contamination, we should thus be able to calculate body burdens directly from the strontium-to-calcium ratio in the urine. A third point I should like to make concerns the power-function. After six years' observation of our strontium cases, we find that the power-function does not fit our excretion data too well, and that at this late stage we get better results using a simple exponential function - with a half-life of about 2000 days. My final point concerns excretion analysis for radium -226. Unless the body burdens are very high, we get too much interference from the normal flow of radium which the person is taking in through his diet, so that in most cases excretion analysis for radium-226 is not satisfactory and we have to resort to measuring the radon content of the breath or whole-body counting. I think excretion analysis for radium-226 is useful only in cases of fresh contamination or in cases with very high body burdens. G.W. DOLPHIN: I think the datayou have mentioned are very interesting. I agree that estimation of retained lung burden from faecal analysis following inhalation is an uncertain procedure. With regard to the strontium-90 excretion following chronic exposure, I should guess that the Sr/Ca ratio in the urine would vary following an acute single intake, owing to the extremely non-uniform bone deposition which would.result from such an intake. H. G. JONES: In section 6.1 of your paper you say that your ’’investi­ gation level" for short-lived nuclides, such as tritium in the form of tritiated water, was one tenth of the maximum permissible body burden. At this level one would not take much action. If, during routine monitoring, workers are found to have about this level of body contamination, continuously, and economic factors make it difficult to low er body burdens much further, would you regard this as a reasonably satisfactory state of affairs? Or would you in fact consider an even higher level satisfactory? G. W. DOLPHIN: I think a higher leve l could be permitted, butit depends on the circumstances. The answers to such questions depend on many factors apart from dose. G. SEEGERS: As you know, there are four radiotoxicity groups, and tritium in the form of tritiated water is placed in the fourth group. . In which toxicity group would you put tritium in the form of an organic compound or a cell compound such as thymidin? G. W. DOLPHIN: I am very much tempted to say I don't know. I think that the whole problem of organic compounds labelled with radioactive ma­ terials is very complicated. I have spent several months trying to answer the problem which has been asked here, and I can give no definite reply. INFLUENCE OF AEROSOL PROPERTIES AND THE . RESPIRATORY PATTERN UPON HAZARDS EVALUATION FOLLOWING INHALATION EXPOSURE

R. G. THOMAS THE LOVELACE FOUNDATION FOR MEDICAL EDUCATION AND RESEARCH, ALBUQUERQUE, N. MEX. , UNITED STATES OF AMERICA

Abstract — Résumé — Аннотация — Resumen

INFLUENCE OF AEROSOL PROPERTIES AND THE RESPIRATORY^PATTERN UPON HAZARDS EVALUATION FOLLOWING INHALATION EXPOSURE. There are three important biological parameters which are necessary in evaluating the hazards from compounds entering the body by any route. These are (1) the amount deposited in-the body; (2) the distribution and translocation kinetics within the body; and (3) the rate of excretion of the material. Sufficient quantitative data on these points are generally lacking in the case of an accidental exposure. This paper deals with experimental animal studies correlating different physical and chem ical charac­ teristics of inhaled particles with the three biological variables mentioned above. Values for* the amount and location of deposited material as a function of the particle size inhaled is presented for a tissue soluble compound (caesium chloride) and for a tissue "insoluble" compound (thorium chloride). Evidence is also given to substantiate the variations which occur in tissue distribution and excretion of an element, depending upon its physical and chemical state when breathed. Data from experiments with aerosols of many compounds, including those already mentioned, is used to show a unique correlation between body burden and faecal ex- . cretion during the first few post-exposure days. The advantages in performing analyses on both urine and faeces for bioassay purposes is demonstrated. The fallacies in current methods of practical hazard assessment from air sampling and bioassay techniques are stressed throughout, using the above data as examples.

INFLUENCE DES CARACTÉRISTIQUES DES AÉROSOLS ET DU RÉGIME DE LA RESPIRATION SUR L'ÉVA­ LUATION DES RISQUES, A LA SUITE D’UNE EXPOSITION PAR INHALATION. Pour évaluer les risques dus aux composés qui pénètrent dans l'organisme par une voie quelconque, il faut tenir compte de trois para­ mètres biologiques importants: 1. Quantité déposée dans l'organisme; 2. Cinétique de la répartition et de la translocation à l'intérieur de l’organisme; 3. Vitesse de l'excrétion de la matière considérée. Dans le cas d'une exposition accidentelle, on manque généralement de données quantitatives suffisantes sur ces para­ m ètres. Le mémoire a trait à des études expérimentales qui ont été faites sur des animaux en vue d'établir une corrélation entre, d'une part, certaines caractéristiques physiques et chimiques des particules inhalées et, d'autre part, les trois paramètres biologiques mentionnés ci-dessus. L'auteur indique des valeurs de la quantité déposée, selon les emplacements et en fonction de la dimension des particules inhalées, pour un composé soluble dans le tissu (chlorure de césium) et pour un autre qui ne l'est pas (chlorure de thorium). Il se base sur certaines données pour expliquer les variations que subissent la répartition tissulaire et l'excrétion d'un élément, selon l'état physique et chimique dans lequel il se trouve lors de l’inhalation. Il se sert de données provenant d'expériences sur des aérosols de nombreux composés, y compris ceux qui ont déjà été mentionnés, pour montrer qu’il existe une corrélation exceptionnelle entre la charge corporelle et l'excrétion dans les matières fécales durant les quelques premiers jours qui suivent l’exposition. Il indique quels sont, aux fins des dosages biologiques, les avantages qu'il y a à analyser à la fois les urines et les matières fécales. En prenant comme exemple les données susmentionnées, l'auteur souligne dans tout le mémoire les erreurs que comportent les méthodes couramment utilisées pour évaluer les risques par les procédés d'échantillon­ nage et d'analyse biologique de l’air.

ВЛИЯНИЕ СВОЙСТВ АЭРОЗОЛЕЙ И ОБРАЗЦА РЕСПИРАТОРА НА ОЦЕНКУ ОПАС­ НОСТИ ОБЛУЧЕНИЯ ПРИ ВДЫХАНИИ РАДИОАКТИВНЫХ ВЕЩ ЕСТВ. Имеется три важных биологических параметров, необходимых для оценки опасности в результате попадания соеди­

3 55 356 R. G. THOMAS

нений в организм любым путем. К ним относятся: 1) количество отложившегося в организме вещества; 2) распределение и кинетика перемещений в организме; 3) скорость выведения ве­ щества. При случайном облучении в целом отсутствуют достаточные количественные данные по этим пунктам. В докладе излагаются данные экспериментальных исследований на животных, показываю­ щие соотношение между различными физическими и химическими характеристиками вдыхаемых частиц и упомянутыми выше тремя биологическими переменными. Представлены данные о количестве и локализации отложившегося материала как функции размера вдыхаемых частиц для растворимого в тканях соединения {хлорид цезия) и для "нерастворимого" в тканях соеди­ нения (хлорид тория). Приведены также доказательства вариаций распределения элемента в тканях и выделения в зависимости от его физико-химического состояния при вдыхании. Данные, полученные в результате проведения экспериментов с аэрозолями многих соединений, в том числе упомянутых выше, использованы для показа уникального соотношения между содержанием соединений в организме и выделением с экскрементами в течение нескольких дней после облучения. Показаны преимущества одновременного проведения для биологических целей анализа мочи и кала. На примере этих данных показаны погрешности, возникающие при использовании при­ меняемых в настоящее время методов практической оценки опасности облучения на основании отбора проб воздуха и методик биологического анализа.

INFLUENCIA DE LAS PROPRIEDADES DE LOS AEROSOLES Y DE SU ESQUEMA REPIRATORIO EN LA EVALUACIÓN DE LOS RIESGOS CONSECUTIVOS A UNA EXPOSICIÓN POR INHALACIÓN. Para evaluar los riesgos derivados de la penetración de sustancias radiactivas en el organismo por una vfa cualquiera, son tres los parámetros biológicos importantes que es preciso conocer: 1) la cantidad depositada en el organismo, 2) la cinética de distribución y translocación dentro del organismo, 3) la velocidad de eliminación de la sustancia por excreción. Por lo general, en el caso de una exposición accidental, se carece de datos cuantita­ tivos suficientes sobre estos puntos. En la memoria se describen los estudios realizados con animales de laboratorio para correlacionar las diferentes características físicas y químicas de las partículas inhaladas y las tres variables biológicas antes mencionadas. Se presentan datos acerca de la cantidad y localización de la sustancia depositada, determinados en función del tamaño de las partículas inhaladas, en el caso de un compuesto soluble en los tejidos (cloruro de cesio) y en el caso de un compuesto insoluble en los tejidos (Cloruro de torio). Asimismo, se facilita infor­ mación encaminada a explicar las variaciones que se producen en la distribución de un elemento en los tejidos y en su excreción, según su forma ffsica y química al ser inhalado. Se exponen datos proporcionados por los experimentos realizados con aerosoles de muchas sustancias, entre ellas las ya mencionadas, para poner de manifiesto la peculiar correlación existente entre la carga corporal y la excreción por vía fecal durante los primeros días consecutivos a la exposición. Se demuestra la ventaja de efectuar a la vez análisis de orina y de heces para determinaciones biológicas. Utilizando los anteriores datos como ejemple, se ponen de relieve los errores inherentes a los actuales métodos de evaluación práctica de riesgos, basados en el análisis de muestras de aire y de sustancias biológicas.

1. INTRODUCTION

This paper deals primarily with factors involved in hazards assess­ ment following inhalation exposure to an aerosol of a potentially toxic sub­ stance. The data to be presented are applicable to both radioactive and non­ radioactive compounds, these latter being commonly referred to as chemi­ cal toxics. Although it is recognized that routes of entry to the body other than inhalation can be of great.importance, for purposes of brevity and unity in presentation, only data from inhalation exposure is discussed. All data to be presented in this paper were obtained from studies on experimental animals, mainly due to the abundance of available information, compared with the scarcity of similar data on humans. It is recognized, and will be emphasized throughout, that quantitative extrapolation from the animal to AEROSOL PROPERTIES AND RESPIRATORY PATTERN 357 man may not be valid, but the qualitative relationships may hold and may give us an insight into the kind of information needed on the human being.

2. TH E G E N E R A L SCHEME

It is obvious that the first step in hazards analysis is a determination (estimation) of the amount of air contaminant retained by the body after breathing the atmosphere for a given period of time. In fact, it is highly desirable to know how the m aterial is distributed along the respiratory tract, whether it is mostly in lung or mostly adhering to the nose, trachea, or other sections of the tract. Once having been deposited, the agent can follow three major pathways, either (a) being raised by ciliary action, swallowed, and excreted in the faeces shortly after exposure; (b) being absorbed into the blood; or (c) remaining in the lung with a certain finite residence time, removal being by either of the aforementioned pathways or by phagocytosis. Of material which is swallowed, a certain fraction may be absorbed into the blood and probably act just as if it came from the lung. Once having been absorbed, regardless of the source, two courses are likely; it can either be deposited in a tissue for a certain period of time, or be excreted in the urine. There are quite good chances that a soluble species in the serum will largely be excreted via the kidney rather than via the bile duct. The rate of loss from the tissues, determined primarily by the chemical nature of the compound, can vary within extremely wide limits and may follow a variety of mathematical relationships.

3. RESPIRATORY TRACT DEPOSITION

The lung model currently used for hazard evaluation by the ICRP [1] has been tested for a long period of tim e and apparently has not led to many striking misjudgements. However, the main reason for this, for radio­ active materials, may well be due to the conservatism which appears to be built into the MPC's (maximum permissible concentration). If there is a conservatism factor of ten in the maximum allowable levels, any m is­ judgement in hazard assessment due to an error in the lung model would probably not be of sufficient magnitude to create a serious problem. The theoretical lung deposition data as calculated by Findeisen and discussed by MORROW may be used to illustrate the last point [2] . His data, which represent only the deposition of particles entering the trachea, and not total- body deposition, show that within the much-quoted "respirable" range (0.05­ 2.0 м т ), the amount which may adhere to the lung tissues can vary by a fac­ tor of « 3. This obviously could affect the estimation of the amount of a toxic agent entering the body, particularly if the only data available for the estimation were derived from an ordinary air sample. Although his data are theoretical, there are numerous experiments which indicate that they may very closely approximate to the actual case. Two such sets of experi­ mental data with albino rats, utilizing aerosols of caesium and thorium, illustrate this. Figure 1 presents the caesium chloride data of LIE, in which animals were exposed to aerosols ranging in equivalent diameters from 0.3 358 R. G. THOMAS

F ig . 1

Total deposition of caesium chloride in the rat as a function of particle diameter to 1.9 цпi [3]. A straight line has been drawn through the points, but it is recognized that this appears linear only because of the lack of data on either end of the size range. The importance of these caesium data is that they do represent 240 rats, and even though they cannot be compared directly to Findeisen's data, they do show the variation that can occur in estimating respiratory tract deposition as a function of particle size. Similar data from the exposure of rats to aerosols of thorium chloride are given in Table I [4]. Here the deposition is separated into Upper R espir­ atory Tract (URT) and Lower Respiratory Tract (LRT), for estimated par­ ticle sizes which vary from a count median diameter of 0.003 pm to 0.1 pm, a factor of over 30. Data for the 0.005 median diam eters d iffer drastically from the others but nevertheless show a trend in the same general direction. If one appraises only the URT deposition, ignoring the two low points, the values are seen to change from 54 at the lowest diameter to 87 at the highest, or only a factor of 87/54 = 1.6. However, in the portion of the resp iratory tract, where the amount deposited may play its most important biological role, namely the lung or LRT, this factor can be seen to increase up to 46/13 = 3.5. Thus these data also show the importance of particle size in estimating inhalation hazards over a completely different range in diameters from the experiments on caesium chloride. It should also be noted that the differences in chemical behaviour in the body between these two compounds (thorium and caesium chloride) are extremely marked, yet they behave si­ milarly in the roles' emphasized here. Certain physiological parameters, such as the tidal volume, minute volume, breathing rate and air-flow rate, can also affect the amount of an inhaled substance which is deposited in the respiratory tract. The one to be discussed here, as an example of their importance, is tidal volume, or, the average volume of air moved during each inspiration or expiration. A person who does not "move" much air with each breath will probably have a fast breathing rate, and it is likely that the air will not have a long resi- AEROSOL PROPERTIES AND RESPIRATORY PATTERN 3 59

TABLE I

RELATIVE DISTRIBUTION OF INHALED THORIUM ONE HOUR AFTER EXPOSURE

Exposure Aerodynamic Total respiratory tractburden No. d iam eter № \

Oim) URT LRT

1 0 .0 0 3 54 4 6

2 0 .0 0 3 54 46

3 0 .0 0 5 9 91

4 0 .0 0 5 16 84

5 0 .0 1 7 64 36

6 0 .0 1 8 53 47

7 0 .0 1 8 . 50 50

8 0 .0 2 7 66 34

. 9 0 .0 6 2 83 17

10 0 .0 7 3 • 74 26

11 0 .1 0 3 87 13

dence time in the alveolar spaces. Conversely, the slow breather with a large tidal volume would be expected to give rise to a longer residence time in the L R T for inhaled particles. 1 Figure 2 shows one method of defining tidal volume as a cumulative log-probit plot. These data were obtained in the studies reported in [3] and represent a total of 50 measurements made for five consecutive days on 10 albino rats. The manner of plotting indicates that tidal volumes in a large population are distributed log-normally. The advantage of being able to make this sort of plot lies in the representative parameters that can be ob­ tained to describe the distribution, namely Mg, the geometric mean and crg, the geometric standard deviation. In Fig. 2 the.Mg is seen to be 1.16 cm3 per breath and the ag = 1.18, indicating an extremely narrow distribution. The importance of tidal volume in body deposition may be seen in Fig. 3 where additional data of Lie, obtained on rats who had inhaled caesium chloride, are presented [3] . The direction (slope) of this curve is the subject of a future publication and should not be considered here, the main reason for this being its tendency to have the opposite slope from that which would be intuitively expected. Nevertheless, its importance for present purposes is to demonstrate the changes in deposition which can occur with tidal volume. These data show a factor of 1.5 in the range of depositions that might have been predicted on the basis of the results of an air sample. Not only is there a distribution of tidal volume values for animals of approximately the same weight (Fig. 2) and a variation in quantity deposited 3 60 R. G. THOMAS

F i g .2

Log-probit plot showing tidal volumes of a "homogeneous" rat population as a function of cum ulative percentage less than the given value

TIDAL VOLUME (cm 3 )

■ . Fig-3

Total-body deposition of caesium chloride as a function of tidal volume in rats

in the body as a function of tidal volume (Fig. 3), but there is also a change in tidal volume with mean body weight and age [5, 6]. With rodents, over a range in weight of 19 to 150 g, there is a change in tidal volume of over a factor of 5, after whichjthere is a sloping off with higher weights. This sloping off is no doubt a function of the attainment of adulthood, the sub­ sequent gain in body weight that ensues, and a decreasing lung-to-body- AEROSOL PROPERTIES AND RESPÍRATORY PATTERN 361 weight ratio while maintaining about the same tidal volume. These same trends occur in human beings [6].

4. TISSUE DISTRIBUTION

Having demonstrated how the amount of an aerosol deposited in the body depends on several factors, the next important facet in the general hazards evaluation scheme is to determine what factors may alter tissue distribution patterns. One good example is contained in Boecker's data, introduced in Table I, in which rats were exposed to thorium chloride aerosols of several mean particle sizes [4]. Figure 4 shows his lung retention values as a func-

' POST-EXPOSURE

Fig. 4

Lung retention curves for four inhalation exposure groups with different initial lung burdens tion of particle diameter, in terms of the amount initially deposited. Note that at 85 d post-exposure there may exist a difference amounting to a factor of over 3 in the "relative" amount of thorium remaining in the lung, depend­ ing upon the particle size. Here the retention time is relatively much lon­ ger for the larger particle sizes which were used in Exposure 9. Converse­ ly, however, the whole-body counting data from these animals, when ex­ pressed as a percentage of the initial lung burden versus time post-exposure, show relative amounts in the whole animal that are almost all in reverse order from the lung curves of the previous graph. The smaller particle sizes (Exposure 1) give rise to a much longer retention time which can amount to a factor of 4 at 60 d. Thus, the larger particles leave the lung at a slower rate at long times after exposure but the early clearance rate is much more pronounced, leading to an overall smaller body burden. The 362 R. G. THOMAS smaller particles appear to be more readily absorbed' from the lung but ap­ parently are véry tenaciously held by other tissues. The skeletal data from these animals are also of interest and particular­ ly in view of the lung and whole-body curves just described. Figure 5 shows

n

F ig . 5

The number of thorium atoms per gram of bone as a function of the initial lung burden that the number of atoms per gram of skeleton varies linearly with the num­ ber of grams of thorium per kilogram body weight initially deposited in the lung [7]. These data show that a fixed fraction of that deposited in the LRT goes to the skeleton, regardless of particle size inhaled, yet the amount deposited in the lung does vary considerably, as shown in Table I. There must be a fraction of the total deposited that remains soluble long enough to gain entrance to the blood and enter the skeleton. That this may be, is backed up by the fact that buildup of activity in the skeleton is very rapid at first with very little or none occurring after a few days post-exposure. Thus, the data may be expressed, as in Fig. 5, as "equilibrium" skeletal values, any changes taking place being quite small after initial fixation [7]. The importance of showing these thorium data lies in the confusion they add to the manner in which distribution patterns may differ as a function of particle size inhaled. With ordinary air sampling procedures, (e.g. , filter paper), and with what present-day quantitative knowledge we have (or do not have) on thorium behaviour within the body, one could easily be off by an order of magnitude in the estimation of body burden either initially or at any time post-exposure. Somewhat similar data are available from inhalation studies with two entirely different (chemically) compounds, na­ mely caesium chloride and niobium chloride [3, 8]. ‘ • One more point may be made on tissue location versus physical or che­ mical properties. Figure 6 cites one example of the differences observed AEROSOL PROPERTIES AND RESPIRATORY PATTERN 363

F ig . 6

Comparison of retention curves from whole-body counting data following inhalation of thorium chloride and thorium citrate '

in relative whole-body retention'of thorium given both as the citrate and chloride, both presumably soluble compounds, [4]. Perhaps this dissimi­ larity in pattern could be expected because of the difference in size of the molecules or the difference in metabolism of the ion versus the chelate. However, the point being emphasized is that sizable variations in body bur­ den can occur but an ordinary air sample would not give an indication of this without a chemical analysis of some sort on thé substance being aero­ solized. It is interesting in this case that after the initial action (transloca­ tion, absorption) both materials led to the same retention kinetics of thorium. All the studies described above involved radioisotopes as tracers for ease of detection. However, the results are directly applicable to ex­ posures involving materials that might be considered chemically toxic. These latter compounds, compared to radioactive toxics, often require larger quan­ tities of the element to manifest their biological effect.

5. FAECAL ANALYSES

In an effort to elaborate on bioassay techniques for this presentation, a review was made of data from faecal analyses performed in many experi­ ments. It was discovered that the easiest and most meaningful way to ana­ lyse these data was to plot the early faecal contents as a ratio of that amount of the compound appearing on a particular day (CF) to the initial body content (C jbc ). If this were accomplished, using both axes as logarithmic, a straight line could be drawn which intercepted the ordinate (CF тС щ ) on day one. It should be emphasized that a log-log plot is very convenient for straightening out many types of biological data, but probably it seldom re­ presents the true function being studied. For convenience of comparison, however, the early faecal data from inhalation exposure to various com­ pounds were plotted this way and a line was fitted by eye, each experimental 364 R. G. THOMAS set of points being fitted independently of the other. In some cases, the points could have been fitted better with a curve, but a straight line was drawn anyway. If a bioassay procedure could be worked out by "eye-ball" fitting, whether it was representing the true function would be unimportant in the practical sense. An example of such a plot is shown in Fig. 7, using

F ig . 7

. Median percentage of initially-deposited antimony appearing in faeces data from the inhalation of antimony in rodents [9]. Similar plots have been made using data from the inhalation of thorium, niobium, polonium and iri­ dium [4, 8, 10, 11, 12] . In some cases more than one experiment was used for the same element; in others, the excretion data from several animals for one experiment were combined because they agreed so closely. In all of these cases, the straight lines closely resembled each other and this can be seen from the presentation of mathematical parameters in Table II. It will be noted that the slopes have a range of from 1.9 to 2.7 and, also, with the exception of iridium, the intercepts fall within quite a narrow range (0.22 to 0.45). Thus, for this wide variety of compounds (and particle sizes) it appears that early faecal analysis might be a very good monitoring tech­ nique to determine the initial body content of an inhaled substance after single exposure. The above procedure appeared very promising, was quite unexpected, and an interpretation of the results was attempted. With a minimum of thought it became quite clear that what is being measured is the emptying time of the gastrointestinal tract, which would be expected to remain quite constant between laboratory animals who have a constant environment (tem­ perature, humidity, food intake). If such a circumstance prevailed in man, this should prove to be a good bioassay technique for estimating initial body burden after a single exposure, within the limits cited. Fortunately (or in this case perhaps unfortunately) man does not exist like the laboratory rat and his living habits are much more varied. However, more data on humans utilizing different aerosols will have to be obtained before the usefulness of this bioassay technique can be ruled out. AEROSOL PROPERTIES AND RESPIRATORY PATTERN 365

TABLE II

EQUATIONS FOR FAECAL EXCRETION CURVES

Antimony chloride С /С = 0.30 t - 2-4 F ШС

Polonium chloride Ср/СШС - 0.45 Г 1-9

Iridium suspension С /С „ = 0. 68 Г 2-2 F ШС

Neptunium dust с /с „ = о. 39 г 2-2 F ШС

Niobium oxalate с д : = о. 37 г 2-7 F ШС

Thorium chloride с /С = 0 . 2 2 г 1-9 F ШС

It should be recognized that for extremely soluble compounds it is not even suggested that this method might work successfully. Also, for daily low-level inhalation exposures, this would appear to have little value.

6. FAECAL TO URINARY RATIOS

Based upon data presented in this paper, even if the-above-suggested faecal analyses were successful as a biossay technique, another measure­ ment would have to determine the fate of the compound remaining in the body. It has been demonstrated how distribution can vary considerably with particle size even though early faecal excretion of the element may not imply this. Thus, some assessment of translocation and solubility within the body fluids is needed. The obvious pathway is an analysis for the element in urine. Urine analysis has been employed as a bioassay technique for years, for many materials, and data on this are plentiful. Since the shortcomings are known, further discussion w ill not be made here. However, the explor­ ation of faecal to urinary ratios, in conjunction with the faecal analyses ela­ borated above, may hold some promise for hazard assessment following single inhalation exposures. Table III shows data collected from dogs after inhaling plutonium dioxide and from rats after inhaling niobium oxalate [13,8]. These ratios were calculated from the mean excreta values of several ani­ mals and were taken from graphs rather than from raw data. They there­ fore should be viewed only for qualitative significance to indicate the types of ratios which can be obtained as a function of air concentration. It will be noted that it is possible to obtain differences in this ratio that can reach a factor of more than ten. This large difference is mostly made up by the amount excrete’d in urine rather than faeces, the urinary excretion following exposure to the smaller sizes of'particles being higher, as expected. Al­ though m ore subtle, the data on niobium show the same pattern, the highest amounts appearing in urine with the smaller particle size. The amounts appearing in faeces fo r these experiments w ere about the same, which also indicates a higher uptake to the internal organs with the smaller particles. Such a scheme as this is mostly applicable to the single exposure case and urine analysis is no doubt still the only feasible solution for the continuous (daily) exposure régime. 366 R. G. THOMAS

TABLE III

FAECAL TO LiRINARY RATIOS AFTER INHALATION

Plutonium dioxide in dogs

Days after exposure Air c o n c .

1 2 3 4 5 7 10 20 30

0 . 5 p c /c m 3 93 56 40 35 2 9 24 20 12 9

1 0 -2 0 122 78 69 60 53 4 4 34 2 6 2 2 '

100 575 500 4 5 6 4 3 6 4 1 4 370 3 3 5 277 2 61

Niobium oxalate in rats

P article Days after exposure size (Mm) 1 2 3 5 7 9 11 15 21 29

0 .2 1 7 .8 1. 7 1 .2 1 . 1 1 .3 0. 94 0 .8 9 0 .9 2 1 .1 0 .6 9

0 .4 3 1 1 .5 3 .1 2 .2 3. 1 4 . 0 - 2 . 9 3 . 0 3 . 5 3 .1 2 . 6

If all of the.above data and related discussions show nothing else, they indicate that for all but perhaps the very soluble and very insoluble com­ pounds, analysis of only one type of excreta is not sufficient to determine the fate of a substance following a single inhalation exposure.

.7. SUMMARY

The* experimental data presented show that a range of greater than a factor of 3 in lower respiratory tract deposition may occur depending upon the size of the inhaled particles. Similar data show a marked change in deposition with tidal volume, and this latter parameter is also shown to vary considerably between individuals. The effect of these variations can lead to errors of greater than a factor of 2 in the assessment of airborne ha­ zards. This same type of erro r, only more pronounced, was shown to appear in tissue distribution as a function of particle diameter. ' A method of body-burden estimation by early faecal analyses following a single inhalation exposure is presented. This has an apparent basis in AEROSOL PROPERTIES AND RESPIRATORY PATTERN 367

the consistent rate of elimination from the gastrointestinal tract under ex­ perimental conditions. The merit in doing both faecal and urinary analyses for estimating body burden is demonstrated. The errors in estimation of inhalation hazards which can occur, unless a thorough air sampling programme is employed, are pointed out.

REFERENCES

[1] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Rep. Int. Sub-Comm ittee II on Permissible Dose for Intern. Radiat. , Brit. J. Radiol. Suppl. 5-7 (1952-57) 23-5 9. [2] MORROW, P. E. , Health Physics 2/4 (1960) 366-378. [3] LIE, R. , MERCER, T .T . and THOMAS, R. G. , Health Physics 9/8 (1963) 878. (Abstract) [4] BOECKER, BRUCE B ., Univ. of Rochester Atomic Energy Project, Rep. UR-605 (1962).

[5] Unpublished data from this Laboratory. . [6] ^Handbook of Respiration, W. B. Saunders Co. , Philadelphia and London (1958).

[7] BOECKER, B.B., THOMAS, R. G. and SCOTT, J.K . , Health Physics 9/2 (1963) 165-176. [8] THOMAS, R. G. , LIE, R. and SCOTT, J. K. , Submitted to the American Industrial Hygiene Association Journal for publication. [9] DJURIC, D ., THOMAS, R.G. and LIE, R ., Univ. of Rochester Atomic Energy Project, Rep. UR-608 ( 1 9 6 2 ) . [10] CASARETT, L. J., Univ. of Rochester Atomic Energy Project, Report UR-552 (1959). [11] BERKE, H.L. and DI PASQUA, А. С., Univ. of Rochester Atomic Energy Project, Report UR-495 (1957).

[12] CASARETT, L. J., BLESS, S., KATZ, R. and SCOTT, J.K ., American Industrial Hygiene Association Journal 21 (1960) 414-418. [13] BAIR, W. J ., WILLARD, D.H. , HERRING, J. P. and GEORGE, L. A. , Presented at the Atomic Energy Commission Inhalation Toxicity M eetings, Richland, Washington, July 5, 6 (1962).

DISCUSSION .

G .W . DOLPHIN: Have you done any histological work on the lung after inhalation? It would be interesting to know what fraction of the particulate matter in the lung was intracellular at various times after inhalation. R. G. THOMAS: We are currently beginning a project on these lines, using insoluble particulates to study clearance mechanisms and auto­ radiography to study cellular localization. B. RAJEWSKY: What is the particle size distribution? R.G. THOMAS: It is very difficult to get homogeneous aerosols having the particle size of interest. Usually the size-frequency distribution is log normal. We are now trying to get insoluble particles which will incorporate the types of isotopes we are interested in. Many people are working on this and I feel sure good results will be obtained. A. KAUL: In Fig. 5 of your paper you have plotted the number of thorium atoms per gram of bone as a function of the initial lung burden. From your results I cannot at the moment calculate the thorium content in bone as a percentage of initial lung burden. What percentage of thorium lung burden is actually translocated to bone? R.G. THOMAS: Approximately 10% of the initial lung burden is trans­ located to the skeleton. ~ 3 68 R. G. THOMAS

60

50

z 30 zэ ex. x 20 ■ t—

10

0 -10 -9 -S -7 -6 -5 - i . -3 -2 -1

LOGARITHM OF DOSE (g T h /k g ) INJECTED

■ Fig. ID '

Percentage deposition of thorium in the skeleton of the rat as dependent upon the mass of thorium injected

A. KAUL: From thorotrast cases we know that only about 0.6% of the total injected thorium is located in the total marrow-free skeleton, while the skeleton containing marrow seems to contain as much as 30%. Did you determine the thorium content in marrow-free bone (compact bone) or are the values related to bones containing bone-marrow? R.G. THOMAS: Yes, marrow was separated from compact bone in some of our studies, but not after inhalation exposure. Following intravenous injection, as in the thorotrast cases, there appears to be a very interesting phenomenon connected with the relative amounts localizing in marrow and compact bone. This may be seen in a graph showing the percentage found in the skeleton as a function of the amount of thorium injected (Fig. ID). At high concentrations there is apparent particulate or colloid formation, pre­ sumably in the injected solution, and the particles are removed by marrow cells of the reticulo-endothelial system. As the injected thorium concen­ tration decreases,. particles become smaller until an ionic form is reached, and the thorium localizes in compact bone as would be expected from chemical considerations. Although this explanation is largely conjectural, it is almost certainly correct. K. BAUMGARTEL: Your paper gives mathematical parameters de­ scribing the faecal excretion curves of various compounds. Is there any correlation between these mathematical parameters and the physical or physiological properties of the compounds examined? R.G. THOMAS: Yes. I am certain that, if the proper experiment were designed to measure the emptying time o‘f the gastro-intestinal tract, the equations would have very much the same parameters as those given. I have not seen any correlation with the physical parameters of the radioisotopes in question, although none of the compounds are particularly soluble in the body. A CRITICAL SURVEY OF THE ANALYSIS ’ OF MICROSCOPIC DISTRIBUTION OF SOME BONE-SEEKING RADIONUCLIDES AND ASSESSMENT OF ABSORBED DOSE

- W. S. S. JEE DIVISION OF RADIOBIOLOGY, DEPARTMENT OF ANATOMY,

Abstract — Résumé — Аннотация — Resumen

A CRITICAL SURVEY OF THE ANALYSIS OF THE MICROSCOPIC DISTRIBUTION OF SOME BONE-SEEKING RADIONUCLIDES AND ASSESSMENT OF ABSORBED DOSE. Detailed m icroscopic dose distribution studies describe the non-uniform distribution of the radiation in contrast to the assumed uniform distribution derived from whole-body retention d ata,. Without some knowledge of the microscopic dose distribution in both space and tim e, it is all too easy to reach incorrect deductions and overlook important conclusions in our attempts to gather comparative toxicity data. Due to the complexity of the problem, the published literature of the measurement of microscopic radiation dose distribution is very limited. The present techniques involve auto­ radiography and are very laborious. They include track counting and microdensitometry. A complete distri­ bution of dose in a given bone in space and time is an immense rask. The microscopic distribution of bone seekers (radium-226, strontium-90, plutonium-239, and radiothorium-228) in the skeleton serves as an ex­ cellent illustration of the complexity of the problem. The initial radiation dose pattern and the altered patterns with time are described for radium and plutonium with particular emphasis on the non-uniformity of the dose distribution in bones and different sites within a bone. For example, the maximum vertebral trabecular surface deposits of plutonium are 1 .5 times hotter than the maximum concentrations on femoral metaphyseal trabeculae. The ratio of dose-rates of surface-deposited plutonium in various sites within the-distal\femur are: metaphyseal trabecular, 3; endosteal, 2.6; epiphyseal trabecular, 1.5; haversian, 1.2; and periosteal, 1. The maximum

localized surface deposit for vertebral trabeculae is 35 to 66 times greater than the calculated average whole- body retention dose-rate. The alteration of the initial non-uniform localization of radium and plutonium with time is discussed in relation to the character of the cellular remodelling of bone and the loss of radio­ nuclide by long-term exchange. Finally, an example using local radiation dose measurements of plutonium and radium with consideration of volume of cells irradiated correlated with the whole-body retention and tumour incidence is used to derive a relative toxicity of plutonium to radium and demonstrate the value of some knowledge of microscopic radiation dose distribution. .

ANALYSE DE LA DISTRIBUTION MICROSCOPIQUE DE'RADIONUCLÉIDES OSTÉOPHÏLES ET DÉTERMI­ NATION DE LA DOSE ABSORBÉE. Des études détaillées sur la distribution m icroscopique des doses perm ettent de décrire la distribution non uniforme des rayonnements qui contraste avec la distribution uniforme supposée que l’on déduit de données sur la rétention dans le corps entier. Sans une certaine connaissance de la distri­ bution microscopique des doses à la fois dans l’espace et dans le temps, on peut très facilement aboutir à des déductions inexactes et négliger d’importantes conclusions dans les travaux tendant à réunir des données comparatives sur la toxicité. En raison de la complexité de ce problème, les ouvrages et articles sur la mesure de la distribution microscopique des doses de rayonnements sont très peu nombreux. Les techniques actuelles font appel à l’autoradiographie et sont très compliquées. Elles comprennent des opérations de comptage de traces et de microdensitométrie. La détermination de la distribution complète des doses à l'intérieur d'un os donné, dans l’espace et dans le temps, constitue une tâche immense. La distribution microscopique des ostéophiles (radium 226, strontium 90, plutonium 239 et thorium 228) dans le squelette constitue une ex­ cellente illustration de la complexité de ce problème. Les auteurs décrivent la distribution initiale des doses de rayonnements et ses modifications ultérieures dans le temps pour le radium et le plutonium en insistant parti­ culièrement sur la non-uniformité de la distribution des doses dans les os et dans les différentes parties d'un os. Par exemple, les dépôts maximums de plutonium sur la surface trabéculaire des vertèbres sont 1, 5 fois plus «chauds» que les concentrations maximums sur les trabécules des métaphyses fémorales. Les rapports des débits de dose des rayonnements émis par le plutonium déposé en surface sur les diverses parties de la zone

distale du fémur sont les suivants: trabécule métaphysaire, 3; période interne, 2 , 6 ; trabécule épiphysaire,

369

24 370 W. S. S. JEE

1,5; canaux de Havers, 1,2; périoste externe, 1. Le dépôt en surface maximum localisé pour les trabécules

des vertèbres est 35 à 66 fois plus important que la moyenne calculée de débit de dose dû à la rétention dans le corps. Les auteurs discutent les modifications de la localisation initiale non uniforme du radium et du plutonium en fonction du temps dans leurs rapports avec la nature du remodelage cellulaire de l’os et avec la perte du radionucléide par échange à long terme. Enfin, dans un exemple avec des mesures de la dose de rayonnements locale du plutonium et du radium, faites compte tenu du volume des cellules irradiées, la rétention du corps et l’incidence des tumeurs sont mises en corrélation et utilisées pour obtenir une toxicité relative du plutonium par rapport au radium et démontrer la valeur d'une certaine connaissance de la distri­ bution microscopique des doses de rayonnements.

~ КРИТИЧЕСКИЙ ОБЗОР ДАННЫХ АНАЛИЗА МИКРОСКОПИЧЕСКОГО РАСПРЕДЕЛЕНИЯ ОТКЛАДЫВАЮЩИХСЯ В КОСТЯХ РАДИОИЗОТОПОВ И ОПРЕДЕЛЕНИЕ ПОГЛОЩЕННОЙ ДОЗЫ. Производилось исследование детального микроскопического распределения"дозы, включающее описание неоднородного распределения радиации, противоречащего принятой тео­ рии однородного распределения, полученного на основании данных задержки изотопов во всем организме. Без некоторого знания микроскопического распределения дозы во времени и пространстве слишком легко сделать неверные выводы и пропустить важные заключения при попытке получить сравнительные данные о токсичности радиоизотопов. Объем опубликованной литературы по измерению микроскопического распределения доз радиации очень ограничен, что связано со сложностью проблемы. Существующие методы включают авто-радиографию и очень трудоемки. Они включают также подсчет распадов и микроденсиметрию. Полное распределение дозы для данной кости во времени и пространстве является огромной задачей. Микроскопическое распределение в скелете откладывающихся в костях изотопов (радий-226, стронций-90, плутоний-239 и радиоактивный торий-228) служит прекрасной иллюстрацией слож­ ности проблемы. Первоначальная доза облучения и доза, изменяющаяся со временем, описаны для радия и плутония, в особенности подчеркивается неоднородность распределения дозы в костях и различных участках внутри кости. Например, максимальные поверхностные трабе­ кулярные отложения плутония в позвонках в 1,5 раза выше, чем максимальные концентрации в трабекулах метафиза бедра. Коэффициент отношения доз поверхностных отложений плуто­ ния в различных участка дистального отдела бедра равен: трабекулы метафиза — 3; эндостиль- 2,6; трабекулы эпифиза — 1,5; гаверсовы каналы — 1,2 и перистиль — t . Максимальные лока­

лизованные поверхностные отложения в трабекулах позвонков в 35 — 6 6 раз выше рассчитанной средней мощности дозы в результате задержки изотопов в организме. Обсуждаются измене­ ния с течением времени начальной неоднородной локализации радия и плутония в отношении характера клеточного воспроизводства кости и потери радиоизотопов при длительном обмене. Наконец, приводится пример измерения местной дозы облучения плутонием и радием с учетом объема облученных клеток и поправкой на общую задержку изотопов в организме и частоту возникновения опухолей для определения относительной токсичности плутония и радия и по­ казывается ценность знания микроскопического распределения дозы облучения.

ESTUDIO CRITICO DE LA DISTRIBUCION MICROSCOPICA DE ALGUNOS RADIONÚCLIDOS OSTEÓFILOS Y EVALUACIÓN DE LA DOSIS ABSORBIDA. Los estudios detallados sobre la distribución microscópica de la dosis proporcionan información sobre la distribución no uniforme de las radiaciones, a diferencia de la distri­ bución supuesta uniforme, que se deduce de los datos relativos a la retención en todo el organismo. En ausencia de cierto conocimiento de la distribución microscópica de la dosis, tanto en el espacio como en el tiempo, se corre el riesgo, al tratar de acopiar datos comparativos sobre la toxicidad, de llegar a deducciones erróneas y pasar por alto conclusiones importantes. A causa de la complejidad del problema, los datos publicados sobre la medición de la distribución microscópica de la dosis de irradiación son muy escasos. Las técnicas actuales recurren a la autorradiograffa y son muy laboriosas. Exigen el recuento de trazas y trabajos de micro- densitometría. La determinación de la distribución completa en el espacio y en el tiempo de una dosis en un tejido óseo dado representa una labor improba. La distribución microscópica de los radionúclidos osteófilos (radio-226, estroncio-90, plutonio-239 y radiotorio-228) en el esqueleto ilustran nítidamente la complejidad del problema. Se describen, con referencia al radio y al plutonio, las características de la dosis inicial de irradiación y las características alteradas con el transcurso del tiempo, insistiéndose particularmente en la falta de uniformidad de la distribución de la dosis en los huesos y en diferentes lugares dentro de’ un tejido óseo determinado. Por ejemplo, los depósitos superficiales máximos de plutonio en la trabécula vertebral son 1,5 veces más activos que las concentraciones máximas en la trabécula de la metáfisis femoral. Las razones entre las dosis provenientes de depósitos superficiales de plutonio en diversos lugares del fémur distal son: trabécula de la metáfisis, 3; endostio, 2,6; trabécula de la epffisis, 1,5; conductos de Havers, 1,2;

24* DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 371

periostio, 1. El depósito superficial máximo localizado en las trabéculas vertebrales es de 35 a 66 v e ce s mayor que la dosis media calculada que corresponde a la cantidad retenida en todo el organismo. La alteración de la localización inicial no uniforme del radio y del plutonio con el tiempo se examina en relación con el carácter de la renovación celular del tejido óseo y con la pérdida del radionúclido por intercambio a largo plazo. Por último, un ejemploi de medición de las dosis locales de irradiación debidas al plutonio y al radio, en el que se tiene en cuenta el volumen de las células irradiadas, que se relaciona con la cantidad retenida en todo el organismo y con la incidencia de tumores, se utiliza para evaluar la toxicidad del plutonio referida al radio y demostrar la utilidad de algunos conocimientos acerca de la distribución microscópica de la dosis de irradiación.

INTRODUCTION

Although it is evident that the distribution of a given radionuclide is often non-uniform,. the knowledge of its non-uniformity is of value in de­ termining the potential hazard of the radionuclide in specific tissues. Most investigators, whose interest lies in bhe assessment of radioactive body burdens in man or experimental animals, will probably not regard detailed investigations of radiation dose distribution as a rewarding research activity. The reluctance of competent investigators to undertake such studies lies not in the fact that they may be uninformed of the value of localized radi­ ation dose studies, but in the complexity of the problem involved and in the knowledge that the techniques for this work are very laborious. Most investigators tend,to believe that these studies represent huge outlays of energy for relatively small scientific returns; thus, few of the current in­ vestigations are actually involved with local dose studies. Those investigators who have been involved in this type of work are often motivated more by their own drive for a greater understanding of the metabolism of a given tissue than for radiation dose studies per se. Bone-seeking radionuclides are known to be unevenly distributed in bone tissues and should serve as a sound illustration of the need for more local radiation dose studies. Thé present report will review previous and current investigations of the distribution of Pu239, Th228, Ra226 , and Sr90 in the ske­ leton, with special emphasis being placed on quantitative studies, of Pu239 and Ra226. These radionuclides serve to describe the two principal types of microscopic distribution in bone and to illustrate the problems in measuring alpha dosimetry in bone. . The problems of calculating beta and alpha dosimetry in bone have been defined lucidly by LAMERTON [1, 2], and he emphasizes how important it is not to underestimate the difficulty of radiation-dose distribution of bone- seeking radionuclides. Bone is a complex tissue; the structure of bone as well as its metabolism and rate of growth and remodelling are known to vary with age and species. The alpha-emitting, bone-seeking radionuclides are troublesome with their short range and their non-uniform distribution in both space and time. Once local radionuclide concentrations,are measured, there still exists the problem of the exact time at which a dose should be integrated and the allowance made for dose-rate, with or without consider­ ation of volume irradiation, before one can assess the absorbed dose. The present discussion will not elucidate or obscure the problem, and the dose will be calculated in both dose-rate and accumulated dose over the burden tim e. 372 W. S. S. JEE

1. CHARACTER OF ADULT BONES

The basis of our radiation measurements involves adult .bones, F igs.l and 2. Adult femurs and vertebral bodies or centra have a shaft (diaphysis) and two ends. These ends are usually wider than the shaft and are known as epiphyses. The part of the shaft adjacent to the epiphysis is wider than

DIAPHYSIS OR EPIPHYSIS

Fig. 1 '

’ Diagram of a typical lumbar vertebral body from an adult Beagle showing that it is similar to a long bone in architecture. It differs only in lacking a true medullary cavity. The dark lines on bone surfaces are the site of bone remodelling, which correspond to sites of radium hotspot deposits. These sites occupied 12>j of the bone surface. the rest of the shaft and is called the metaphysis. In the adult, the bone tissue of the metaphysis and the epiphysis is continuous and consists of ir ­ regular anastomotic trabeculae which form a type of bone known as spongy or cancellous bone. The spaces between the trabeculae are filled with bone cells, marrow cells, and other primitive cell types. The external parts of the epiphysis and metaphysis consist of a thin layer of compact bone. The shaft of the femur is a tube of compact bone, the cavity of which is known as a medullary (marrow) cavity. The diaphyseal portion of the ver­ tebral bodies consists of a tube of compact bone, but lacks a true medullary cavity, and, instead, is filled with trabeculae and marrow cells. The outer surface of bones is' surrounded by a connective tissue sheath of periosteum, while the inner surface is lined by a thin cellular layer of endosteum.

2. THE CHARACTERISTICS OF BONE SURFACES

There is common agreement that tumours tend to arise from the osteo­ genic tissue at bone surfaces and a basic understanding of the cellular con­ stituents related to these bone surfaces is needed for our discussion of radi- DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 373

SPONGY BONE (TRABECULAE)

PERIOSTEUM

ENDOSTEUM

CORTICAL 0ONE (HAVERSIAN SYSTEM OR OSTEON) DIAPHYSIS (SHAFT)

■ Fig-2

CHARACTER OF LONG BONE (DISTAL FEMUR) Diagram of a typical distal femur of an adult Beagle

, FI0ROCYTE UNOIf FERIENTIATEO (ON RESTING SURFACE) MESENCHYMAL CELL

Fig- 3 ■

Diagram of a typical bone surface of a portion of a trabecula showing osteoblasts, osteoclasts and fibrocytes lining growing, resorbing and quiescent bone surfaces. Only the undifferentiated mesenchymal cells (reticular cells) have been depicted as occupying the marrow spaces. Note the pre-osseous layer (osteoid) beneath the osteoblasts. 374 W. S. S. TEE

ation dose distribution and absorbed dose. The character of bone surfaces has been thoroughly discussed recently by other investigators [3, 4]. Bone surfaces are found on trabecular bone, Haversian canals, or on periosteal or endosteal surfaces. These surfaces can be readily divided into three types: growing, quiescent, and resorbing. These three types are best shown on a trabecula, Fig. 3. Growing surfaces are characterized by the presence of osteoblasts apposed to a thin layer of uncalcified matrix (osteoid). The uncalcified matrix is a secretory product of osteoblasts. The quiescent surfaces are characterized by endosteal cells or fibrocytes (resting osteo­ blasts) lining the bone; while the resorbing surfaces are characterized by irregular bone surfaces lined by giant multinucleated cells, the osteoclasts.

3. MICROSCOPIC DISTRIBUTION OF Pu239 AND Th228

The pattern of distribution of Th22a and Pu239 in bone is quite sim ilar [5-13]. The initial deposits of these radionuclides are on all mineralized bone surfaces: periosteal, endosteal, resorption cavities, forming osteons, Volkmann and Haversian.canals, Fig. 4. The surface deposits of radio­ nuclides are uneven, and the mechanism determining their concentrations is unproven [4, 5, 14, 15]. Bone remodelling alters the initial surface de­ posits by bone resorption and apposition. As a result of this remodelling process, osteoclasts and macrophages contain concentrations of radio­ activity, and a low-level volume of plutonium or radiothorium is deposited in post-injection bone, Fig. 5. The low er amount of activity in post-injection bone is determined by the rate of bone apposition and the concentration of radionuclide in the plasma. Therefore, continual remodelling results in an increased volume of bone receiving alpha irradiation; more prolonged burden times result in more uniform distribution of plutonium or radiothorium in bone, Fig.5. In reality, however, uniform radionuclide deposition in cor­ tical bone is never achieved. The turnover of bone by cellular processes (remodelling) averages approximately 2%/yr in cortical bones of both adult human [16, 17] and adult I3eagle tibia [18]. The replacement process is also non-uniform in that it may remodel one site more than once without disturbing another. Thus, the probability of a uniform deposition in bone can never be achieved under 50 yr post-contamination. Nevertheless, uniform distribution in trabecular bone is achieved in less time, and in lumbar ver­ tebral bodies of adult Beagles, most of the trabeculae are observed to con­ tain a uniform distribution of plutonium after five years post-injection.

4. LOCAL RADIATION DOSE STUDIES OF Pu239

All published radiation dose-distribution studies dealing with this radio­ nuclide are from adult Beagle studies performed in Utah. Our studies have quantitated the non-uniform deposition of Pu239 after a single intravenous injection. In these studies, the non-uniform distribution of plutonium is compared to the uniform label. The uniform label represents the plutonium concentration per mass of the bone (skeletal burden) that would exist if this radioelement were uniformly distributed throughout the skeleton. Body DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 375

SURFACE DEPOSIT

. F i g .4

Diagram showing the localization of Pu239 or T h 228 on bone surfaces. Note that some of the osteocytes are out of range of the alpha rays.

Fig. 5

Diagram showing the burial of heavy Pu239 concentrations located on growing bone surface and

the removal of Pu239 and bone by osteoclasts on the apposing surface. The osteoclasts concentrate

plutonium, while the Pu239 localized on quiescent bone surfaces remains on the surface. The recently-formed bone contains a volume deposit of plutonium. ■ burdens of plutonium can be determined by whole-body counting, chemical analysis of bone, and excretion studies. Since 10-30% of the plutonium may 376 W. S. S. JEE concentrate in the liver, an intrinsic error exists in calculating local dose from whole-body counting and excretion studies. Therefore, radionuclide content of the skeleton is best determined by chemical ashing of bones and subsequent counting of these preparations for radioactivity [19, 20]. The non-uniform bone surface deposits of plutonium in lumbar vertebrae and distal femurs have been determined by using track counting and m icro­ densitometry of quantitative autoradiograms. . . Unlike the alkaline earths, the uptake of plutonium and radiothorium is initially restricted to bone surface and this deposition is unaffected by age [21]. However, their concentrations differ from bone to bone and in sites within a given bone. The trabecular surface deposits in lumbar verte­ bral bodies are 1.5 times higher than the trabecular surface deposits in the femoral metaphysis of the same dog. In the distal femur, the concentration of bone surface deposits are, respectively: metaphyseal trabecular, 3; en­ dosteal, 2.6; epiphyseal trabecular, 1.5; Haversian (osteonic), 1.2; and periosteal, 1. Moreover, the surface deposits within a site are uneven. In this respect, Fig. 6 shows the frequency distribution of Pu239 concentrated along central trabecular surfaces of a lumbar vertebral body from a dog sacrificed 28 dpost-injection. In this bone, the deposits varied from 0.37 to 6.75 pc Pu239/cm2, and the average concentration equalled 0.0239±0.117pc Pu239/cm2 [22].

5. - ASSESSMENT OF ABSORBED DOSE FROM PLUTONIUM

Table I summarizes the absorbed dose average over the first Ю дт of osteogenic lining cells apposed to bone surfaces in adult Beagles injected with a single intravenous dose of three different dose levels. Dr. C.W. Mays in our laboratory has derived the formula for calculating the absorbed dose [23]:. Some typically measured and calculated maximum and average bone surface dose-rates in lumbar vertebral centra are listed along with the uni­ form label dose-rates calculated from skeletal retention data. Ratios of the maximum and average dose-rates to uniform dose-rates vary between 10 and 86. These non-uniformity factors are exceedingly large in that the ratio of maximum surface concentration is about 76, and the average sur­ face concentration to uniform label is 18. The post-injection volume de­ posits resulting from burial of the initial surface deposits by bone concen­ trating less radioactivity are calculated to deliver about 4.4 rad/d to the osteogenic cells within 10 jum in the dogs injected with 2.7 цс Pu239 /kg, Fig. 7. This dose-rate is similar to the average skeletal dose-rate (uniform label = 4.7 rad/d), but is 22 times less active in the local surface concentration of Pu239 and emphasizes the impact of remodelling processes in decentra­ lizing the intense initial location of the radionuclide.

6. MICROSCOPIC DISTRIBUTION OF Ra22<¡ AND Sr90

The other group of bone-seeking radionuclides contains the alkaline earth elements, radium and strontium, and these radionuclides deposit in sites identical to calcium in bone. These radioelements deposit in high con- DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 377

s u r f û c e : d e p o s i t DIFFUSE DEPOSIT

Fig. 6

Diagram showing the complete removal of part of the intense bone surface deposits resulting in a volume deposit in part of the trabecula, while the undisturbed, quiescent surface remains the same. In the case of radium, the entire trabecula would contain a diffuse concentration (see text).

FREQUENCY DISTRIBUTION OF Pu” 9 CONCENTRATION ON ENDOSTEAL SURFACES OF A DOG.(TI2P5)

2 239 10 XjUjuc Pu PER UNIT AREA

Fig.4

Frequency distribution of plutonium on bone surfaces of a lumbar vertebral body from an adult Beagle injected with 2. 74 pc Pu23Vkg and sacrificed after 28 d 37 8 W. S. S. JEE

HOTSPOT OH BONE SURFACE

Fig. 8

Diagram showing the distribution of radium hotspots in the region of new bone growth and the diffuse distribution in remaining old bone

Fig. 9

Diagram showing the burial of radium hotspots by the addition of new bone containing less radioactivity. ' The quiescent surfaces remain undisturbed with the diffuse concentrations. centration in rapidly-calcifying bone matrix beneath osteoblastic surfaces and osteoid as "hotspots" (the growing surfaces) and throughout the pre­ existing old bone in areas of lighter, more uniform concentration called the "diffuse" component, Fig. 8 [10, 12, 24-35]. It-is important to realize DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 379

. TABLE 1 . .

TYPICAL MEASURED AND CALCULATED DOSE-RATES ADJACENT TO BONE SURFACES IN LUMBAR VERTEBRAL CENTRA OF ADULT BEAGLES ADMINISTERED SINGLE I V INJECTION OF PLUTONIUM

Average dose-rate to first 10 jim

Injected dose M axim um A verage . (M c/kg) M áxim u m A verage Uniform Uniform Uniform surface surface label (ra d /d ) (rad /d ) (ra d /d )

2 .7 278 9 9 .0 4 . 7 60 21

0 . 3 3 6 .7 5 . 3 0 . 5 . 73 10

0 .0 1 5 2 . 6 0 .7 0. 03 86 23

that hotspots form only where active bone growth is occurring, and this same active apposition of bone, which is essential for radium and strontium hotspot concentration, is the very process which promptly buries the hot­ spots within less radioactive bone, Fig. 9. Bone formed after injection is diffusely labelled and generally contains less radium than bone formed before injection, and the amount of radium deposited in this post-injection bone is proportional to the specific activity of the plasma at the time of bone ap­ position. Thus, these extremely high concentrations of radium in hotspots are removed from the bone surface. In both cortical and trabecular bone, the hotspots are buried and can be removed by normal or pathological re­ modelling; however, some buried hotspots do persist throughout the life­ span of the animal or individual. At the time of death, after many years of contamination with radium, hotspots represent the new bone growth that took place during the exposure period, while the diffuse deposit represents old bone present at the time of injection incorporating its radium by ex­ change and secondary mineralization, and post-injection new bone growth. Thus, the quiescent bone surface contains a diffuse radium deposit. These surfaces make up approximately 90% of the available bone surfaces.

7. LOCAL RADIATION DOSE STUDIES OF RADIUM

M ore detailed local dosim etry on the deposition of radium in the skeleton- is available than for any other bone seeking element. The apparent reason is that clinical materials are available and there is a great deal of financial support for research in this area because radium - as measured in man - is used as a standard for comparing the toxicity of other bone-seeking radio­ nuclides. The Oxford group has analysed terminal bones from two dial painters with body burdens of radium [33, 34], but the bulk of the information collected, both in quantity and quality, has been performed by ROWLAND at Argonne, Illinois [16, 28, 32, 38-41]. A study is also being carried on 380 W. S. S. JEE in our laboratory on radium deposition in the skeleton of dogs subjected to chronic toxicity of the radionuclide element. These dogs are serially sacri­ ficed after being injected intravenously with a single dose of one of six dose levels of the radionuclide. Although the aim of this review is to emphasize the non-uniformity and the removal of radium detected by the analyses of local distribution. of radium, it must be emphasized that a recent report of Rowland [32] beautifully illustrates the information that can be deduced from terminal distribution of radium in human bone. This information includes the rate of acquisition of the radium, the route of administration, the magnitude of each dose, and the terminal body burdens. These types of information are invaluable in that radium contamination in man is very poorly documented. In addition, terminal body burdens have been calculated by the English group from alpha track counting of term inal bone specimens [33, 34]. There are many approaches to expressing the non-uniformity ratios for the microscopic distribution of radium. It can be expressed as a ratio of either hotspot to diffuse, hotspot to uniform label, or diffuse to uniform label. The uniform label is determined by whole-body counting and is the body burden divided by the skeletal mass. In our laboratory, hotspot measurements have been restricted realistically to the maximum hotspot reading. In the past, hotspot values have been measured by a technique which involves the determination of a frequency distribution of hotspots. This has been abandoned because it was found to be too time-consuming.

8. RADIUM MEASUREMENTS IN MAN

In Tablé II are listed some typical measurements and calculated non­ uniformity ratios from Rowland's study of 18 cases [16]. This investigator uses the ratio of specific activity of the hotspots to the adjacent diffuse dis­ tributions as a measure of the degree of non-uniformity. More significant is a non-uniformity ratio of maximum hotspot to the uniform label, a value which can be compared to other bone-seeking radionuclides. This com­ position will give a value which will describe the maximal local uptake of radium as compared to a body burden assuming uniform label of the radio­ nuclide. The diffuse to the uniform value has been reported to be of greater bio­ logical importance [3, 35]. M ARSHALL [35] has reported that the accumu­ lated dose from radium hotspots is not more than a few times the dose due to the diffuse, owing to the self-burial of hotspots with less radioactive bone, and only about .10% of the bone surface is involved in bone apposition. There­ fore, only 10% of the bone surfaces initially contain hotspots, which are rendered less radioactive immediately by bone apposition. The remaining bone surfaces have the diffuse concentration of radium. Thus, 90% of the soft tissue adjacent to bone surfaces is irradiated by a radiation dose equi­ valent to the diffuse concentration, Fig. 1, and 10% of the bone surfaces are temporarily irradiated with a higher dose-rate. In man, Rowland [16] has observed the ratio of hotspot to diffuse aver­ ages 73 for 18 cases, and the values range widely from 16 to 218. The cases exposed to radium for a shorter period than the chronic exposure via the DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 381

TABLE И

NON-UNIFORMITY RATIO IN RADIUM PATIENTS [16]

Ratio T erm in al M ethod o f Patients radium burden acquisition M axim u m hotspot M axim u m hotspot Diffuse (MC) Diffuse Uniform Uniform

FR 7 .0 dial painter 94 30 0 .3 2

313 1 . 3 dial painter 19 7 . 3 0 .5 0

0 1 . 2 ' I V medically 150 7 0 ,6 0 .4 7

I 0 . 8 I V m ed ically 2 18 8 6 .7 0 .4 0

watch dial industry show much higher hotspot to diffuse ratios. The specimen with a ratio of 218 was a patient who received only 18 intravenous injections medically. When radium is present in blood for a relatively short time (medical intravenous injection) the resulting hotspot to diffuse label is high. With longer exposure (dial painters), the ratio is lower. The ratio of diffuse to uniform label averages 0.5 in the 16 cases, with a range of 0.23 to 0.90, while the ratio of maximum hotspot to uniform averages 30 with a range of 7.0 to 86.7.

9. COMPARATIVE RADIUM MEASUREMENTS

In Table III the various ratios in compact bone of adult Beagles are compared to the human data. The bones involved are the femoral and tibial shafts and the ratio of hotspot to diffuse is about 20, with a ratio of diffuse to uniform of 0.6, and a ratio of hotspot to uniform of 12 [36]. Surprisingly, the diffuse to uniform ratios in both species are similar, while the hotspot to diffuse ratios differ by at least a factor of 3. In man, the hotspot to dif­ fuse ratio has been reported as high as 263 for 10 weekly injections of 10 цс Ra226 [32], as compared to 20 for a single intravenous injection of radium in young adult Beagles. This is an obvious example of species variation in the avidity of radium to mineral of new bone of human and dog.

10. RADIUM MEASUREMENT IN ADULT BEAGLES

In the Utah studies [36] of radium in dogs, the non-uniformity values of various bones and different sites within the same bone were measured (Table IV). In these bones, the ratios of maximum hotspot to diffuse and maximum hotspot to uniform label are quite similar, but there is an obvious difference in the uptake of radium in old trabecular bone. The diffuse to uniform ratio of spongy bone differs significantly from that of cortical bone. In this respect, the spongy bone possesses a diffuse to uniform label of about 382 W. S. S. JEE

TABLE III

NON-UNIFORMITY RATIO IN CORTICAL BONES OF RADIUM PATIENTS AND ADULT BEAGLES

M ethod of Maximum hotspot Maximum hotspot Diffuse acquisition Diffuse Uniform Uniform

Radium patient M ixed 7 3 . 1 1 6 1 . 9 . 29.2 ±10. 8 0. 47 ± 0. 23 (Rowland)

Utah Beagles

Tibial shaft 1 I V inj. 2 1 . 6 ± 5 .9 1 2 .5 ± 4 .5 0. 55 ± 0. 06

Femoral shaft 1 I V inj. 2 0 . 3 i 4 .7 12.4 ± 2.3 . 0 .6 3 ± 0 .0 2

TABLE IV

NON-UNIFORM LOCALIZATIONS OF RADIUM IN CORTICAL AND SPONGY BONES IN ADULT BEAGLES

T ype o f Maximum hotspot Maximum hotspot Diffuse Bone bone Diffuse Uniform Uniform

T ib ia C o m p act 21.6 ± 5.9 12.5 ± 4.5 0 . 55 ± 0. 6

Fem ur C o m p act 20.3 ± 4,7 . 12.4 ±2.3 0.63 ±0.02

Lum bar Spongy 18.8 ± 4.7 21.3 ±6.2 1.16±0.11

Femur (M) ^ Spongy 13.4 ±4.8 13.8 ±5.4 1. 03 ±0.16

Femur (E) (b) Spongy . 1 9 . 1 ± 5 .3 1 4 .1 ± 4 . 9 0 .7 8 ± 0 .0 3

Femoral metaphysis. (b) Femoral epiphysis (old spongy bone).

one, while that of the cortical bone is about 0.6. In other words, the diffuse deposits in these areas of spongy bone are equal to the uniform label derived from whole-body retention studies. Whether this value of one is comparable to the diffuse to uniform ratio in spongy bone of man remains to be seen.

11. LOSS OF RADIUM DETECTED BY QUANTITATIVE MEASUREMENTS

Before any accurate estimate of accumulated dose can be calculated, it is necessary to describe and quantitate the loss of radium with time. Loss of radium from the skeleton of man and adult Beagles has been described by a power function of the form Rt = atb where Rt is the fractional retention DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 383

at any time t_ (in days) after acquisition of radium, and a is the fractional retention at one day, while the coefficient b has a value of -0.52 in man [37] and -0.2 [38] in adult Beagles. The value (b) described in the above power function includes the loss of radium by cellular resorption and ex­ change. The rate of loss of radium by these processes has been quantitated by distribution studies [16, 36, 39-42]. Resorption is the process by which a given volume of bone mineral and matrix is removed from bone. This process normally is related to the internal reconstruction or remodelling of bone, whereby radium can be removed into body fluids and some of it relocated in the bone, with the net result that less radium is picked up by the new bone. The second process, known as exchange, is any process in­ volvin g equal and opposite rates of tra n sfer of atoms to and from a single volume of bone mineral. Cellular remodelling can drastically reduce a local concentration of radium if the event involves the replacement of a hotspot or a portion of hotspots with bone containing less radioactivity. The remodelling rate for man is about 1-2%/yr in cortical bone [16] and has been shown to vary with age [17], while remodelling rates for spongy bone are yet to be characterized. The radium loss by exchange has been characterized and quantitated by local measurements of concentration in a unique experiment by Rowland [42]. He compared the unremodelled hotspot to adjacent diffuse radium con­ centrations in two radii at two tim e intervals from the same adult dog. One radius is removed at 4 weeks and the other radius is removed one year later at sacrifice. The bone removed at sacrifice contained 74 ±5% as much Ra226 /g as the one removed a year before by amputation. The radioactivity in hotspot concentration decreased to 66 ±12% of the 4-week level and the diffuse decreased by 75± 15% of the original level. SEARS et al. [36] at Utah have also measured loss of radium by using a power function to describe the retention of radium in hotspot and diffuse deposits unaltered by resorption in serial sacrificed dogs. Table V lists the regression coefficients (b) for maximum hotspot and average diffuse radium deposits in spongy and cortical bone tissue of various bóne and em­ phasis that, within the same bone, the regression coefficients are similar, with the only exception being the coefficient for the diffuse deposits in the femoral epiphysis.. Similar coefficients exist for vertebrae and mandibles, but these coefficients differ from those obtained for the tibia and femur by a factor of two. Surprisingly, the rate of loss is similar in both compact and spongy bone of the distal fem ur and the difference appears to be between bones rather than types of bones (spongy and compact). The data also in­ dicate that some bones (mandibles and vertebrae) lose radium by exchange more rapidly than the whole-body retention value of -0.2, while some bones (femurs and tibiae) lose radium by exchange at a slower rate than for the entire skeleton.

12. C A L C U L A T IO N OF ABSORBED DOSE

Dose-rates are calculated from local radium measurements and ter­ minal radon retention values from the equation derived by C..W. Mays in our laboratory [23]: The calculation is for the average dose-rate to the 384 W. S. S. JEE

TABLE V

REGRESSION COEFFICIENT (b) FOR RADIUM DEPOSITS IN ADULT BEAGLES INJECTED WITH 10 pc Ra226/kg [36]

A verage Bone Type of bone Maximum hotspot Diffuse

Lumbar vertebrae Spongy - 0 .2 6 - 0 .3 0

Femoral metaphysis Spongy - 0 . 073 - 0 .1 3

Femoral epiphysis Spongy - 0 .1 0 - 0. 014

Femoral shaft C o rtica l - 0 .1 4 - 0 .1 3

Tibial shaft C o rtical - 0 .1 3 - 0 .1 1

Mandible (compact) C o rtical - 0 .2 6 - 0 .2 6

(a ) W hole body reten tion = y = 0. 079 t '° '2 first 10 /jm of soft tissue adjacent to a bone surface. The accumulated dose is simply the terminal dose-rate times the burden time and it does not con­ sider the loss of radiation by exchange. It is not a true accumulated dose, but it*does represent the minimum accumulated lifetime dose. The dose- rates and accumulated dose from hotspot, diffuse, and uniform labels are listed at three different time intervals for adult Beagles injected with a single 10 /ис Ra226/kg (Table VI). In this respect, dose values are quite variable. If the maximal dose is considered the critical dose, the resulting values are many factors higher than the average dose received by the skeleton (uniform label). If the diffuse dose is considered the critical dose, then these values are much low er and may be many factors below the average dose to the skeleton. Terminal dose-rates and minimum lifetime dose to a 10 цт lacuna (osteocyte) have been calculated for the human data by Rowland (16) in Table VII. These are terminal dose values and probably are many orders of magnitude below the dose-rates immediately after exposure to radium.

13. QUANTITATIVE STUDIES WITH Sr90

The bulk of the excellent quantitative radiation dose studies are from VAUGHAN and OWEN at Oxford [30, 44-48]. Their values were derived from experiments with rabbits at various ages, subjected to various du­ rations of exposure to Sr90 . In addition, there is an extensive monograph by the Swedish workers [31] and one report from Hanford from miniature swine which were fed Sr90 daily [49]. The Utah group has a joint project with the Oxford group; however, at this time, I am not at liberty to mention any of the data obtained from this current investigation. The bulk of the literature on quantitative Sr90 has been well-documented and only the findings on non-uniform concentration will be received. The tragedy of the local dose measurement of Sr90 in rabbits lies in the fact that whole-body retention measurements (uniform label) are lacking. ITIUIN F OESEIG AINCIE 385 8 3 RADIONUCLIDES BONErSEEKING OF DISTRIBUTION

' TABLE VI

DOSE CALCULATIONS FOR ADULT BEAGLES INJECTED WITH 10 дс Ra2^6/kg

Uniform label L. vertebral hotspot L. vertebral diffuse T o ta diffuse

Av. rad to skeleton Av. dose to 1st 10 цт Av. dose to 1st 10 /im A v. dose to 1st 10 jjm Burden time (d ) ra d /d A cc . rad - ra d /d A c c . rad ra d /d A c c . rad - ra d /d A c c . rad

12 19 2 3 0 1 6 4 1 9 6 8 1 1 . 3 . 1 3 6 3 . 1 ‘ 3 7 -

9 0 8 1 1 . 3 1 0 4 0 0 8 5 . 8 7 7 9 0 0 5 . 7 5 1 7 6 2 . 0 1 8 1 6

1 3 8 0 1 1 . 3 1 5 5 0 0 7 5 . 4 1 0 4 0 5 2 5 . 3 7 3 1 4 2 . 0 ■ 2 .7 6 0 386 . W. S.'S. JEE

TABLE VII

TERMINAL DOSE CALCULATION FOR RADIUM PATIENTS

Terminal dose*rate Minimum lifetime

to a 10 ¿im lacuna dose to a 10 pm lacuna T erm in al Patients body burden (PC) Hotspot Diffuse Hotspot Diffuse ..(rad /d ) (rad /d ) (rad /d ) ■ (rad /d )

FR 7. 0- 2 4 .0 0 .2 5 240 000 C500

313 1 .3 2 .2 0 .1 2 2 4 000 1300

Q 1 .2 9 .5 0 .0 6 4 90 000 610

T 0 . 8 1 0 .0 0 .0 4 8 95 000 460

In order to be able to compare the degree of non-uniformity of Sr90 to other radionuclides, it is desirable to express non-uniformity as the ratio between the maximum or minimum concentrations to the uniform label as measured by whole-body counting. A ratio of non-uniformity expressed as the ratio of maximum-to-minimum concentrations is meaningless in the case of plu­ tonium, where the value would be infinity. In the initial deposition pattern of plutonium, there are areas of bone and marrow devoid of any radiation; therefore, the minimum value is zero. A non-uniform ratio of maximum- to-minimum concentration of radium also will result in higher values, es­ pecially when a hotspot deposit is compared to a deposit of radium in a re­ cently-formed osteon. The non-uniformity ratios reported to date for rabbits are listed in Table VIII. These values represent the ratios of maximum and minimum dose after the injection of 600 цс of Sr90/kg to weanling and adult rabbits. These ratios emphasize the variation in individual bones, and the effect of age and time after injection on radionuclide content of bone. These non­ uniformity ratios (maximum-to-minimum dose) vary from 2 to 20 [48]. Age

TABLE VIII

VARIATION IN RATIO OF MAXIMUM-TO-MINIMUM DOSE TO WEANLINGS AND ADULT RABBITS INJECTED WITH 600 pc Sr90/kg

W eanlings Adults

Bone

1 day 6 months 1 day 4 6 months

Jaw 5 ■ 2 10 20

T ib ia 16 5 10 10

V ertebra 5 3 7 1 10 DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 387 also appears to be very important as the maximum-to-minimum ratios differed significantly for a weanling as compared to an adult animal. In assessing the dose to the vertebra after an injection of 600 ц с /kg of S r 9o, the absorbed dose, was 360 rad/d (first dfay) and was reduced drastically to 60 rad/d at 180 d. ' ENGSTROM et al. [31] in their monograph on bone and radiostrontium, described a ratio of Sr90 between the mineral content of old bone and newly- formed Haversian systems to be 1:10 or even 1:20. Most frequently the ra tio of la b ellin g was about 1:5 between new and old bone. ■' Vaughan and Owen [44] further describe a ratio of maximum-to-minimum concentration of Sr90 0f 4 in continuously-fed rabbits; while at Hanford [49] the ratio reached 8 in miniature swine continuously fed Sr90 . Also, in­ valuable information on Beagles continuously-fed Sr90 should be forthcoming soon from a project at Davis, California, United States.

14. CONSIDERATION OF VOLUME IX DOSE CALC U LATIO N

Due to the short range of the alpha-emitting, bone-seeking radionuclides one must involve the consideration of volume irradiated when assessing the absorbed dose from Pu23®, Th 22S; and Ra226 . One can calculate the volume of tissue at risk by simply multiplying the endosteal surface of a bone by the range of the alpha or beta particles. If one takes the total endosteal surface area of 4100 mm2 of a typical dog thoracic vertebral body, the marrow volume to be 644 mm3 [50], the critical tissue as the osteogenic tissue at bone surface, the most critical cell assumed to be the more pri­ mitive cell types in the marrow, and the range of Pu239 series (35.4 цт ), R a ‘226 series (66.2 цт ), Th228 series (82.1 цт ) and Sr90 s e ries (1.2 cm) in soft tissue, one will find the following per cent of their marrow is within range of their radiation: Pu239 - 22%; Ra226 - 41%; T h 228 - 67%; and Sr90 - 100%. Such variation in the volume at risk-should be taken into consideration in calculating comparative dose values. • . There have been attempts to resolve the problem of the consideration of volume irradiated in the assessment of absorbed dose. These include: (1) the concept of the interval between successive passage or spacing of ionizing particle disintegration through a 10 jum.diam. cell (5 цт diam. nucleus) with each alpha track transversal through a 5 цт diam. nucleus involving about 100 rad; and (2) the concept of the mean dose delivered to a given distance from the bone which will involve the entire range of the alpha track. . Lamerton [2] first introduced the concept of the interval between the_ successive passage of ionizing particles through à 10 цт diam. cell with a 5 цт diam. nucleiis and that the mean dose per nucleus per passage would be equivalent to 100 rad. He calculated that the passage of an alpha through any part of a cell occurs about once every 56 yr for a given nucleus for a uniform tissue distribution of Pu239 giving 30 mrad per week.' A similar treatment of'our dose calculation for plutonium in adult Beagles is listed in Table IX. It lists the disintegration of alpha tracks per day in microns along the bone surface for the three dose levels. At 99 rad/d, one.alpha track per day disintegrates at an interval of 2.8 д т. In other words, every 388 W. S. S. JEE

TABLE IX

ALPHA TRACK DISINTEGRATIONS PER DAY ALONG BONE SURFACES CONTAMINATED WITH Pu239'

Maximum dose-rate Average dose-rate . Injected dose. M axim um , M inimum 1 st 10 (jm 1 st 10 jjm ' . (M c/kg) (rad /d ) (i'm ) ' (rad /d ) (№)

2 . 7 278 0 .1 9 9 .0 2 . 8

. 0 . 3 ■ 3 6 .7 1 . 6 5 .3 53

0 .0 1 5 2 . 6 110 0. 7 400 nucleus (assuming 5 цт. diam. nucleus) of all cells lining bone surfaces is traversed by two alpha particles every day when there is a dose-rate of 99 rad/d to the first.10 д т of soft tissue. At the dose-rate 0.7 rad/d the passage of an alpha track is a rare event with an incidence of one track per day at a distance of 400 цт between cells or a maximum one hit in every 40 cells lining bone surfaces. Unfortunately, the, above concept does not consider the range of the plutonium alpha particle and the non-uniform dis­ tribution of energy along'its 35.4 yum path in soft tissue. The other approach is to calculate the mean dose-rate for the entire range of the alpha track. Unfortunately the range of Pu239, Th228 and its dàughters, and Ra226 and its daughters d iffe r, and consequently the radiation dose from the bone surfaces are different. SPIERS [51] has shown that the radiation dose decreases as the distance from the radiation increases. Also, C. W. Mays has calculated a most useful table which breaks down the mean do'se-rates at 10 цт intervals delivered by a surface (Pu239 and Th228) or volume (Ra226) deposits of 1 nc ( 10-3 /лс)/ст2 or cm3 of the three radio­ nuclides. These dose-rates are plotted >in Fig. 10. It serves to demonstrate the diminishing dose-rate at increased distances from bone surfaces and the’ impossibility of comparing dose-rates delivered by radionuclides of dif­ ferent ranges at a given site. For instance, it would be impossible to com­ pare the mean d ose-rate d elivered by Pu 239 and Th228 and its daughters at 50 to 60 цш from the bone surface. Th22s and its daughters deliver 7.1 rad/d and the Pu239, zero. Although the dose-rate delivered by Th228 deposits on bone surface at 50 to 60 цгп is small, it may be sufficient to induce some late biological effects. Further experimentation is needed to clarify the biological consequences of such a small dose-rate for cells located in this region from the bone surface. Meanwhile, it may be wise to consider the mean radiation dose over 80 цт. fo r Ra226, Pu239 , and Th 228.

CONCLUDING REMARKS ■ '

It is obvious from the discussion of microscopic distribution arid the attempted assessment of absorbed dose from bone seekers that it is not suf­ ficient to characterize a radioactive burden by describing the total content DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 389

. F i g . 1 0

Plot of dose-rate from surface deposits of Pu 239 series and Th228 series and volume deposits of

Ra226 series of 1 ne (10 -3 /j c ) /c m 2 or cm3. (By courtesy of Dr. C. W, Ways. ) of the skeleton. If bones are available for analysis, local radiation dose studies should be undertaken and whole-body retention studies should serve as a point of departure. Conversely, those investigations describing the distribution of local radiation doses should determine whole-body retention values as a basis for comparing non-uniform concentrations, and, if possible, the two approaches should be inseparable. The limitation in the autoradio­ graphic method of measuring dose-rate is the fact that measurement can be made only at one point in time. Any estimates of accumulated dose re­ ceived by the soft tissues must involve extrapolation, using serial sacrifice animals or information on loss of radionuclide during the animal life-span by whole-body counting. Investigations of the distribution of local radiation dose of Pu239 , Ra226 and Sr90 have contributed somewhat to our understanding of the non-uniformity factors in a few species. The non-uniformity factors have been described to vary as follows: (a) types of radionuclide; (b) rate of acquisition; (c) age; (d) condition of. skeleton; (e) different bones; (f) sites within bone; (g) types of bone; and (h) species. The major shortcoming is the real lack of in­ formation on the measurements of local concentrations of these bone-seeking radionuclides. Only the non-uniform distribution of radium can be described for man. There are insufficient data to extrapolate the non-uniformity factor of Sr90 and Th228 for man from animal experimentation. It appears that our 390 W. S. S. JEE

best bet now and in the future is for more information about the local dis­ tribution of radiation of bone-seeking radionuclides coming from dog studies. Finally, the radiation dose delivered by bone-seeking radionuclides cannot be considered as a purely physical problem; the biological aspect of bone tissue (metabolism of bone) must also be considered. It has been mentioned for the calcium-like radionuclide, radium, that the loss of this element by cellular remodelling and exchange' must be characterized before it is possible to determine the radiation dose. The fate of the actinide ele­ ments, Pu239 and Th228 f which are foreign to osseous tissue, are also con­ trolled by cellular remodelling. Furthermore, the studies of local radiation dose distribution coupled with a better understanding of bone metabolism should provide us with the proper selection of the radiation concentration to calculate absorbed doses as well as the critical tissue. For instance, it has been described from microscopic distribution studies that plutonium deposits on bone surfaces, and that the surfaces of bones are particularly sensitive to the induction of bone tumours. Also, it is often not the highest concentration of radionuclide which is effective in bone tumour induction. At present it is unreasonable to relate the potential toxicity of radium to the maximum concentration (hotspot). Hotspots are usually buried within bonë and their alpha rays are not irradiating the more sensitive tissues on bone surfaces, while the less-concentrated diffuse component occupying 90% of the available bone surfaces is more effective in osteosarcoma induction. Indeed, this is supported by the data from the calculation of the comparative effectiveness of plutonium relative to radium in producing osteosarcoma, using the data from the Utah study on the chronic toxicities of Pu23® and Ra226 in adult Beagles [52] and assuming that the absorbed dose to the first 10/um of cells adjacent to bone surface is the critical dose. It has been observed that adult Beagles injected with 10 цс Ra226/kg and 0.9 /uc-Pu239/kg induced osteosarcomas in approximately 1500 d. If the relative effectiveness, is based on injected doses, then plutonium is 11 times more effective than radium. If the maximum concentrations deliver the effective dose, then Ra226 is 1.8 times more effective than plutonium. But, if the diffuse de­ posits on bone surfaces are the critical dose, then the plutonium is eight times more effective than radium. It is more reasonable to conclude that the diffuse deposition of radium should be the critical radiation dose and that plutonium is eight times more effective than radium in producing osteo­ sarcoma. . This conclusion is supported by detailed autoradiographic studies showing the burial of hotspots and their removal from bone surfaces. This example also serves to demonstrate that local radiation doses can be mis­ leading unless there is abetter understanding of bone metabolism. Of course, our calculations have been simplified by ignoring the important consider­ ation of volume irradiated. If the consideration of volume irradiated is in­ cluded in our calculation, then Pu239 effectiveness to Ra226 should be reduced by almost a factor of two to 4.2.

• ACKNOWLEDGEMENT

This paper contains a summary of the work done, or being done, by a number of former and present colleagues. I would like to express my grati- DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 391 tude particularly to D r. J. A. Twente, Mr. N.L. Dockum, Dr. K.A. Sears, M rs. R. M ical and M r. R .K . Haslam, with special thanks to Dr. C. W..Mays, whose equations made it possible to calculate our dose-rates. I am grate­ ful to Miss J. Clayton for her help in preparing the manuscript. Finally, I wish to express my appreciation to Professor T.F. Dougherty and the United States Atomic Energy Commission for their continual support and encouragement. . .

REFERENCES

[1] LAMERTON, L. F. , in: Radioisotopes in the Biosphere (C ald eco tt, R. S. and Snyder, L. A ., Eds. ) Univ. of Minn. , Minneapolis (1960) 382-400. [2] LAMERTON, L. F. , in: Some Aspects of Internal Irradiation (Dougherty, T. F. , Jee, W. S. S. , Mays, C. W. and Stover, B. J. , Eds. ) Pergamon Press, Oxford (1962) 397-403. L3] ARNOLD, J. S. , in: Therapy of Radioelement Poisoning (Rosenthal, M. W. , Ed. ) Argonne National Laboratory, ANL-5584 (1956) 131-43. ■

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[10] ARNOLD, J. S. and Jee, W.,S. S. , Lab. Invest. 8 (1959) 194-204. , [11] JEE, W. S. S. and ARNOLD, J. S. , Lab. Invest. 10 (1961) 797-825. ' [1 2 ] JEE, W. S. S . , ARNOLD, J. S. , COCHRAN, T. H. , TW ENTE, J. A. and MICAL, R S. , in: Som e Aspects of Internal Irradiation (Dougherty, T. F ., Jee, W. S. S ., Mays, C. W. and Stover, B. J ., Eds. ) Pergamon Press, Oxford (1962) 27-45.

[13] ARNOLD, J. S. and JEE, W. S. S. , Hlth Phys. 8 (1962) 705-07.

[1 4 ] JEE, W. S. S. and ARNOLD, J. S. , Hlth Phys. 8 (1962) 709-11. '

[15] FOREMAN. H. , Hlth Phys. 8 (1962) 713-16. [16] ROWLAND, R. E. , in: Radioisotopes in the Biosphere (C a ld e co tt, R. S. and Snyder, L. A. , Eds. )U niv. of Minn., Minneapolis (1960) 339-53. [17] FROST, H. M. , C añ ad. J. B iochem . Physiol. 41 (1 9 6 3 ) 3 1 -4 2 . [18] JEE, W. S. S. , in: Bone B iodynam ics (Frost, H. , Ed. ) L ittle Brown and C o . , Boston (1 9 6 4 ) [19] STOVER. B .J.. ATHERTON. D. R. and KELLER. N.. Rad. Res. 10 C19591 130-47. [20] STOVER, B.J. , Hlth Phys. 1 (1959) 373-78. . "

[21] SCHUBERT, J. , Nucleonics 8 (1951) 13-28. , [22] TWENTE, J. A ., BUTLER, E. G. , FREUDENBERGER, O. and JEE, W. S. S. , in: Radiobiology Laboratory Progress Report COO-218 (Stover, C. N. , Jr. , Ed. ) Univ. of Utah, Salt Lake City (1959) 190-206. [23] MAYS, C. W. and SEARS, K. A. , in: Research in Radiobiology, COO-226 (Dougherty, T. F ., Ed. ) Univ. of Utah, Salt Lake City (1962) 78-85. ' [24] HOECKER, F. E. and ROOFE, P. C ., Radiology 56 (1951) 89-91. [25] ARNOLD, J. S. , Amer. J. Physiol. 167 (1951) 765. [26] AUB, J .C . , EVANS, R. D. , HEMPELMAN, L. H. and MARTLAND, H. S. , M ed icin e 31 (1 9 5 2 ) 2 2 1 -3 2 9 . [27]. LOONEY, W. B. and WOODRUFF, L. A. , AMA Arch, of Path. 56 (1953) 1-12. [28] ROWLAND, RE. and MARSHALL, J. H. , Rad. Res. 11 (1959) 299-313. - [29] KIDMAN, B., TUTT, M. L. and VAUGHAN, J. , J. Path, and Bact. 62 (1950) 209-227. ' [30] VAUGHAN, J. , in: The Biochemistry and Physiology of Bone (Bourne, C. G. , Ed. ) Academic Press, New York (1956) 729-65. . . [31] ENGSTROM, A ., BJORNERSTEDT, R ., CLEMEDSON, D. J. and NELSON, A. , Bone and Radiostrontium, John Wiley and Sons, Inc., New York (1958) 80-98. [32] ROWLAND, R. E., in: Diagnosis and Treatment of Radioactive Poisoning, IAEA, Vienna (1963) 57-68. [33] HINDMARSH, М ., OWEN, М ., VAUGHAN, J. , LAMERTON, L. F. and SPIERS, F. W. , Brit. J. Radiol. 31 ' (1958) 518-33. . 392 W. S. S. JEE

[34] HINDMARSH, М ., OWEN, М.. and VAUGHAN, J. , Brit. - J: Radiol! 32 (1 9 5 9 ) 1 8 3 -8 7 . ■ [35]- MARSHALL, J. H. , in: Radioisotopes and Bone (McLean, I.e ., Lacroix, P. and Budy, A ., Eds.) . Blackwell, Oxford (1962) 35-46. - ' [36] SEARS, K. A ., JEE, W. S. S . , HASLAM, R. K. and MAYS, C . W. , in: Research in Radiobiology, C O O -227 (Dougherty, T. F. , Ed. ) Univ. of Utah, Salt Lake City (1963) 90-106. . [37] NORRIS, W.P.., SPECKMAN, T. W. and GUSTAFSON, P. F. , Amer. J. Roentgenol. Rad. Therapy and nucl. Med. 73 (1955) 785-802.

[38] VAN D ILLA,~M . A ., STOVER, B. J. , FLO YD, R. L. , ATHERTON, D. R. and TA YSU M , D. , Rad. Res. 8 (1958) 417-37. • [39] MARSHALL, J. H. , ROWLAND, R. E. and JOWSEY, J. , Rad. Res. 10 (1959) 258-98. [40] MARSHALL, J. H. , in: Bone as a Tissue (Rodahl, K. , Nicholson, J. T. and Brown, E. M. , Jr., Eds. ) Blackiston D iv., McGraw-Hill (1960) 144-55. [41] ROWLAND, R. E. , Clin. Orthoped. 17 (1960) 146-53. • [42] ROWLAND, R E ., Rad. Res. 15 (1961) 126-37. [43] FROST, H. M. , Bone Remodelling Dynamics, Charles C. Thomas, Springfield (1963) 102-08.

[44] VAUGHAN, J. and OWEN, М ., Lab. Invest. 8 (1959) 181-93. ' [45] OWEN, M. , SISSONS, H. A. and VAUGHAN, J., Brit. J. Cancer 11 (1957) ,229-48. [46] OWEN, M. and VAUGHAN, J. , Brit. J. RadioL 32 (1959) 714-2Ü . [47] OWEN, M. , and VAUGHAN, J. , Brit. J. Cancer 13 (1959) 424-38. [48] OWEN, М ., in: Some Aspects of Internal Irradiation (Dougherty, T. F. , Jee, W. S. S. , Mays, C. W. and Stover, B .J., Eds.) Pergamon Press, Oxford ( Í 962) 409-20.

[49] M cCLELLAN, R. O. . BUSTAD, L. К. , CLARKE, W. J. , DOCKUM, N. L. , McKENNEY, J. R. and KORNBERG, H. À ., in: Some Aspects of Internal Irradiation (Dougherty, T. F ., Jee, W. S. S. , Mays, С. W. and Stover, B. J. , Eds. ) Pergamon Press, Oxford (1962) 341-48. [50] TWENTE, J. A. and JEE, W .S.S., HIth Phys. 5 (1961) 142-48. [51] SPIERS, F. W. , Brit. J. Radiol.. 26 (1953) 296-301. [52] DOUGHERTY, T. F. , STOVER, J. H. , DOUGHERTY, J. H. , JEE, W. S. S. , MAYS, C. W. , REHFELD, С. E . , .CHRISTENSEN, W. R. and GOLDTHORPE, H. C. , Rad. Res. . r 7 (1 9 6 2 ) 6 2 5 -8 1 .

DISCUSSION

R.G. THOMAS: I don't understand how the dose was calculated in the marrow from a surface deposit of the radioactive material. Was any account taken of maxima of ionization density along the alpha track? . W.S.S. JEE: The maximum would be at the surface. The dose-rates were calculated for intervals of energy absorption of 0-10/nm, 10-20/jm, 20-30|um etc. away from bone surface. We are aware of the difference of ionization density along the alpha track but have not included it in the calcu­ lation, so as to facilitate understanding of the concepts involved. ' B. RAJEWSKY: What is your definition of hotspot? W.S.S. JEE: My definition of hotspot is quite simple. It is a site of new bone formation. At these sites new minerals are being created, and this is where the highest concentration of material generally accumulates. B. RAJEWSKY: And what is the approximate size of your hotspots? W. S. S. JEE: They vary. They can be 300 or 400/um on the bone surface of a trabecular. Now, on an average about 10% of the bone surface in the vertebral body is occupied by hotspots and the.rest is not. So 10% of the bone surface in the vertebral body is remodelling and forming new bone, and this coincides with the hotspot distribution. . B. RAJEWSKY: And what of the tissue in the intervals between the hotspots? The diffuse concentration? DISTRIBUTION OF BONE-SEEKING RADIONUCLIDES 393

W. S. S. JEE: It is just sitting there waiting to be remodelled eventually. Some of it may sit through one's life-span. B. RAJEW SKY: But what is the behaviour of the tissue in this interval between two hotspots? W. S. S. JEE: W ell, at these high doses which we are dealing with in the dog we think the diffuse deposition is the site where the bone tumour is formed. In the hotspot, we believe that these high concentrations kill the osteocytes and the bone-forming and bone-lining cells. B. RAJEWSKY: Causing necrosis? W .S.S. JEE: Yes - at my dose level. At a lower dose-level the hotspot may be the critical site. , B. RAJEWSKY: I agree that cancer form ation is bound up with the d if­ fuse dose. I never saw tumours in the hotspots but often at a distance of 20-100 /um from them. ' I do not, however, understand why you take a ratio of hotspot-to-diffuse dose. These are two quite different biological situations. In my view the doses are not comparable, and their ratio is meaningless. W.S.S. JEE: The only reason I have stuck to a hotspot-to-diffuse ratio is that it has been used before, and I just wanted to conform to practice. I thought I would be attacked at this particular point about the significance of hotspots. Everyone is placing too much emphasis on the hotspots. John Marshall has shown that by definition a hotspot is a site of new bone formation. That means this particular site will be buried by bone, and this is what we generally see in dogs. I think Marshall sees it with human and mouse material too. It is buried, and once you bury this hotspot, you no longer have a dose that is attacking the soft tissues, so the important thing is the diffuse dose. I myself would rather use the maximum hotspot - to-uniform ratio, but the most important ratio certainly is the diffuse- to-uniform ratio. And we do still need to find out the diffuse-to-uniform ratio for spongy bone in man. For cortical bone the diffuse-to-uniform ratio is almost the same in the dog, but it's higher in dog for spongy bone. Per­ haps we could discuss this further in private. ■ C. J. MALETSKOS: I would agree with your comments that hotspots are not or may not be important in bone tumour formation on condition you can prove that osteogenic sarcomas do not or cannot arise from osteocytes. W.S.S. JEE: I must apologise if I have given the impression that we should ignore the osteocytes as a critical cell in the formation of osteogenic sarcomas. In the past few years there was fair agreement that the bone-lining cells and their stem cells were the critical cells, but less than two years ago Professor Bélanger and his associates showed the active role of osteocytes in the osteolysis of bone. I believe, therefore, the osteocytes should be brought back into the picture as a potential critical cell. We are in fact back where we started.

SYMPOSIUM ON THE ASSESSMENT OF RADIOACTIVE BODY BURDENS IN MAN

HELD AT HEIDELBERG, 11-16 M A Y 1964

CHAIRMEN OF SESSIONS Sessions 1 and 2 W. V. MAYNEORD Institute of Cancer Research, University of London, . England

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