CONFERENCE PROCEEDINGS OF THE 21ST RADIATION HYGIENE DAYS JASNÁ POD CHOPKOM, SLOVAKIA, 23 - 27 NOVEMBER, 1998
SK98K0328
RADIATION HYGIENE DAYS 1998
ERRATUM
CONTENS: THE SYSTEM FOR AUTOMATIC DOSE RATE MEASUREMENTS 154 BY MOBILE GROUPS IN FIELD D. Drábová, R. Filgas, I. Češpírová, M.Ejemová
AUTHOR INDEX: Češpírová I. 135,154 Ejemová M. 154 Drábová D. 154 Filgas R. 154
Published by Nuclear Regulatory Authority of Slovak Republic Edited by Helena Cabaneková, Matej Ďurčík, Denisa Nikodémova Printed in Bratislava by GRAFIT3 Number of Copies 180
ISBN 80-88806-04-6
30-16 SOCIETY OF NUCLEAR MEDICINE AND RADIATION HYGIENE OF SLOVAK MEDICAL ASSOCIATION, BRATISLAVA INSTITUTE OF PREVENTIVE AND CLINICAL MEDICINE, BRATISLAVA NUCLEAR REGULATORYAUTHORITY OFSR, BRATISLAVA SLOVAK ENERGETIC ENTERPRISES
MEMBERS OF ORGANISATION COMMITTEE Dana Drábová, Ing. ,CSc. Peter Gaál, MUDr. Izabela Makaiová, Doc, MUDr., CSc. Denisa Nikodémova, RNDr., CSc. František Parimucha, RNDr. Zdeněk Prouza, Ing., CSc. Marián Štubňa, Ing., CSc. Milan Vladár, Ing., CSc. Jozef Zlatňanský, Ing., CSc.
MEMBERS OF SCIENTIFIC COMMITTEE Helena Cabáneková, RNDr. Matej Ďurčík, RNDr.
CONFERENCE SPONSORS CANBERRA PACKARD, s.r.o. WIENERBERGER, Slovenské tehelne, s.r.o. POWER PLUS, s.r.o. BAUMITHIROCEM, s.r.o. SLOVENSKÉ ELEKTRÁRNE, a.s. BRAMAC - strešné systémy, s.r.o. HEINEKEN, Slovensko, a.s. SERMAN s.r.o.
THE EDITIONAL BOARD IS NOT RESPONSIBLE FOR THE CONTENT OF PAPERS, AS WELL AS FOR UNCORECT ENGLISH TRANSLATION 21" RHD Jasná pod Chopkom
PREFACE
The Society of Nuclear Medicine and Radiation Hygiene of Slovak Medical Association, a Member of International Radiation Protection Association, together with the Institute of Preventive and Clinical Medicine in Bratislava, submits to you the Proceedings of the 21th Radiation Hygiene Days, organized in Jasná pod Chopkom (Low Tatras) from November 23 to 27,1998.
The main conference topics are focused on current problems in Radiation Protectioa Thanks to the great effort of ICRP, IAEA and other international organizations, the basic principles of radiation protection are set and probably would not be changed in the near future. Though the fact, that these basic principles are adopted, the priorities have to be set with regarding the primary aims of radiation protection, with applying the appropriate standards of protection and safety at the national levels. Here is great effort to attain the standards of the more developed countries. It is known, that all measures taken against any hazard depends on the financial resources which should cover the protective measures.
In addition to other political and social values, the radiation protection must correspond to widely accepted ethical values of the people affected by these regulations. Ethical values relating to protection require that the total harm to human health should be minimized and substantial resources should be allocated to the protection of human health.
In the Slovak regulations, the radiation protection philosophy and policies of Basic Safety Standards which are based on the three concepts of justification, optimization and limitation, as well as on the general requirements for practices and interventions, are reflected. What is still needed is the system of operational documents and decrees in the field of radiation protection. These documents should be in agreement with the legislation of the European Communities. Nevertheles the organization of the radiation protection shoud continue to be under the state control..
The today economic situation in our country does not allow to cover all protective measures. We hope at least in continuing process of the proper training and qualification of professionals, together with the possibility to be equipped with the appropriate devices and systems. For reviewing and assessing the effectiveness of the protective measures the quality assurance programmes should be included.
The most important source of information on risk in human are comprised in epidemiological studies. In Slovakia some basic information, necessary for assessment of the health effects of various radiation sources on the population are already gained. With regard to the assessment of possible risk, it is necessary to continue this research work. It is challenge for us to connect these studies with highly appreciated Slovak cancer epidemology register, which is one of the best in Europe. 4 2ľRHDJasnápodChopkom
The papers presented in the Proceedings are not edited, also the tables and figures are not changed.
We would like to thank all members of the organizing committee for excelent work during the conference preparation, mainly the two editors : RNDr. Helena Cabáneková and RNDr. Matej Ďurčík.
We are grateful for the help and support of the Slovak Medical Association, prof. MUDr. Tomáš Trnovec, Dr.Sc, director of IPCM, Slovak Nuclear Regulatory Authority and Slovak Electric,a.s.. Finely I would like to thank all sponsoring organizations and exhibitors.
Denisa Nikodémova Vicepresident of the IRPA associated Slovak Soc. of Nuclear Medicine and Radiation Hygien 21st RHD Jasná pod Chopkom
CONTENS
PREFACE 3
CONTENS 5
WHAT ARE THE PURPOSES OF RADIATION PROTECTION 11 M. Vladár
STATE SUPERVISION OVER RADIATION PROTECTION IN THE 16 CZECH REPUBLIC Z. Prouza
RESULTS OF RADIATION PROTECTION AT NUCLEAR POWER 20 PLANT DUKOVANY 1988-1997 Z. Zelenka
ANALYSES OF OCCUPATIONAL DOSES AT THE DUKOVANY 24 NPP J. Filip, E. Fiala, B.Jurochová, Z.Zelenka
MONITOROVANIE AKTIVITY OSÔB A VOZIDIEL NA VÝSTUPE 28 Z SE-EBO JASLOVSKÉ BOHUNICE Ľ.Dobiš, J.Kaizer, J. Svitek
QA REQUIREMENTS FOR CZECH NUCLEAR ACTIVITIES AND 33 RADIATION PRACTICES D. Jarchovský
INFORMACE O SPRÁVĚ ÚLOŽIŠŤ RADIOAKTIVNÍCH ODPADŮ 36 V. Starostová
WHAT IS MORE DANGEROUS: NUCLEAR POWER PLANTS OR 37 CARBON FIRED POWER PLANTS ? J. Kuruc
THE NUCLEAR REGULATORY AUTHORITY OF THE SLOVAK 42 REPUBLIC -INFORMATION TO THE PUBLIC M. Šeliga
PREČO A AKO KOMUNIKOVAŤ S VEREJNOSŤOU O 47 RADIAČNOM RIZIKU S.Ftáčniková
IMPROVEMENT OF INFORMATION ON THE NUCLEAR ENERGY 52 HEALTH EFFECTS, THE AIM OF WIN SLOVAKIA M. Petrášová, D. Nikodémova 6 21" RHD Jasná pod Chopkom
RADIONUCLIDE COMPOSITION OF NOBLE GASES IN 56 EFFLUENTS OF SOME NUCLEAR ESTABLISHMENTS P. Rulik, M. Tomášek, I. Malátová
RADIOTRACER STUDY OF THE ADSORPTION OF Fe(HI), Cr(III) 59 AND Cd(II) ON NATURALAND CHEMICALY MODIFIED SLOVAK ZEOLITE M. Fôldesová, P. Dillinger, P. Lukáč
THE STUDY OF THE UPTAKE OF PLUTONIUM BY SLOVAK 61 ZEOLITES P.Lukáč, N.Patzeltová, M.Foldesova, P.Dillinger
THE DEVELOPMENT OF NATIONAL REGISTRATION SYSTEMS 63 OF RADIATION PROTECTION IN THE CZECH REPUBLIC K. Petrová, Z. Prouza
EVALUATION OF IRRADIATION IN PATIENT'S ENVIRONMENT 67 RECEIVING 131I THERAPY J.Husár, A.Furiová, F.Borovičková
RADIATION LOAD OF POPULATION DUE TO TREATMENT OF 72 CHILDREN NEUROBLASTOMA BY 131I-MIBG J.Heřmanská, J.Zimák, P.Kavan, P.Došel
ABOUT CALCULATION OF THE EQUIVALENT DOSE FOR 76 AUGER EMITTERS USED IN DIAGNOSTIC NUCLEAR MEDICINE S.Ftáčniková
INITIAL EXPERIENCE WITH SOFTWARE SYSTEM JODNEW FOR 82 EVALUATION BIOPHYSICAL CHARACTERISTICS RELATED TO TREATMENT OF CARCINOMA OF THYROID GLAND BY mI J. Heřmanská, J. Zimák, L. Jirsa, M. Kárny, K. Vošmiková, J. Nimec, T. Blažek
3 D GEL DOSIMETRY-METHOD REVIEW AND OUR FIRST 87 EXPERIENCE J.Novotný Jr., V.Spěváček, J.Novotný, T.Čechák, J.Vymazal, J.Tintěra
RADIODIAGNOSTIC MEASUREMENTS RADIATION LOAD OF 92 CHILDREN BY CHEST D.Nikodemová, M.Ranogajec, M. Vladár, M.Horváthová
RADIODIAGNOSTIC MEASUREMENTS RADIATION LOAD OF 96 CHILDREN BY CHEST D.Nikodemová, M.Ranogajec, M. Vladár, M.Horváthová 2 ľ RHD Jasná pod Chopkom
RADON EXPOSURE AND LUNG CANCER RISK 97 L.Tomáäek
RESULTS OF CYTOGENETIC EXAMINATIONS OF MINERS 102 EXPOSED TO RADON IN ORE MINES M.Beňo, M.Vladár, D.Nikodemová, M.Vičanová
ANALYSIS OF THE SIMILARITY FACTORS OF THE VILLAGES 107 IN THE AREAS OF THE NUCLEAR POWER PLANTS FROM THE PREMATURE DEATH-RATE PERFORMED BY FUZZY LOGIC METHOD M. Letkovičová, R. Rehák, J. Korec, B. Mihály, V. Príkazský
TLD AUDIT IN RADIOTHERAPY IN THE CZECH REPUBLIC 113 D.Kroutilíková, H.Žáčková, L.Judas
RADIATION PROTECTION PROBLEMS BY THE OPERATION OF 121 THE CYCLOTRON FACILITY M.Ďurčík, D.Nikodemová
VERIFICATION OF NUCLEAR MEDICINE RADIONUCLUDE 125 CALIBRATORS IN SR A.Švec
RECENT STAGE OF THE EVALUATION OF MEDICAL 128 EXPOSURES IN THE CZECH REPUBLIC AND DIAGNOSTIC RADIOPHARMACEUTICAL DOSE ESTIMATE TO THE CZECH POPULATION K. Petrová, V. Husák
INDIVIDUAL DOSIMETRY IN HIGH ENERGY RADIATION 130 FIELDS F.Spurný
MONITORING OF THE INTERNAL CONTAMINATION IN THE 135 CZECH REPUBLIC. SURVEY FOR THE NEEDS OF EURADOS GROUP I. Češpírová, I. Malátová
CASES OF OLD INTERNAL CONTAMINATION WITH241 Am 139 LMalátová, Š.Foltánová, V.Bečková
TISSUE FREE WATER TRITIUM SEPARATION FROM 140 FOODSTUFES BY AZEOTROPIC DISTILLATION F.Constantin, A.Ciubotaru, D.Popa 2 ľ RHD Jasná pod Chopkom
THE AIR CONTAMINATION BY l37Cs AND 7Be IN THE 144 TERRITORY OF SLOVAK REPUBLIC H.Cabáneková, M. Vladár
7Be IN AMBIENT ATMOSPHERE AND IN INDOOR AIR 149 F.Ďurec, A.Ďurecová, Ľ.Auxtová, E.Gombala
RADIOCHEMICAL ANALYSIS OF ENVIRONMENTAL SAMPLES 152 P.Dillinger, M.Harangozó, J.Tolgyessy
INTERCOM? ARISON MEASUREMENT - ORAVA SOIL 156 F.Ďurec, A.Ďurecová, Ľ.Auxtová, E.Gombala
IN SITU GAMMA RAY COUNTING OF THE LARGE VOLUME 159 OBJECTS USING THE LARGE AREA PLASTIC SCINTILATORS Š.Krnáč
PERSONAL DOSIMETRY SERVICE IN THE SLOVAK REPUBLIC 164 J.Compel
METHODS OF ASSESSMENT OF WHOLE BODY241 Am CONTENT 167 Š.Foltánová, I.Malátová, J.Klisák
CONTAMINATION OF A NEUTRON GENERATOR FACILITY BY 171 TRITIUM II. M.Tomášek
CONTRIBUTION TO THE PENETRATION OF RADIONUCLIDES 174 ACROSS THE SKIN. CONCENTRATION DEPENDENCE OF 60Co PERMEATION Z.Kassai, V.Koprda, M.Harangozó
FINDS OF RADIOACTIVE MATERIALS IN CENTRAL SLOVAKIA 179 IN 1996-1998 Ľ.Auxtová, F.Ďurec, P.Adámek
AIRBORNE GAMMA RAY SPECTROMETRY - METHODS, DATA 183 PROCESSING AND APPLICATIONS L.Klusoň, T.Čechák, P.Jurza
A COMPARISON OF SEMICONDUCTOR GAMMA 187 SPECTROMETRIC ANALYSIS USING THE PEAK NET AREA CALCULATIONS AND THE WHOLE SPECTRUM PROCESSING Š.Krnáč, M.Koskelo, R.Venkatamaran 2 ľ' RHD Jasná pod Chopkom
PLASTIC SCINTILATOR GUARD GATES FOR MONITORING THE 192 RADIOACTIVITY OF THE RAILWAY WAGONS Š.Krnáč
THE UNINTENTIONAL AND UNCONSCIOUS EXPOSURE TO 197 RADON (AND OTHER NATURAL RADIONUCLIDES) J.Thomas
METROLOGY OF RADON AND THORON CONCENTRATIONS 201 M.Ďurčflc, M.Vičanová
NATURAL RADIOACIVITY IN SLOVAK CONSTRUCTION 205 MATERIALS AND THE INDOORS DOSE RATE FROM BUILDING MATERIALS H.Cabáneková, M. Vladár
RADIATION LOAD FROM RADON EXPOSURE IN SLOVAKIA 209 M.Vičanová, M.Ďurčík, D.Nikodemová
222Rn CONCENTRATION IN THE OUTDOOR ATMOSPHERE AND 213 ITS RELATION TO THE ATMOSPHERIC STABILITY K.Holý, R.Bôhm, I.Bosá, A.Polášková, O.Holá
CONTINUAL MONITORING OF RADON DECAY PRODUCTS 217 CONCENTRATIONS IN INDOOR AND OUTDOOR AIR P.Petruf, K.Holý, T.Stanys
RADON IN WORKPLACES - APPLICATION OF NEW SLOVAK 222 LEGISLATION M.Futas, E.Gombala
RADON EMANATION COEFFICIENTS IN SANDY SOILS 225 K.Holý, A.Polášková, A.Baranová, O.Holá, I.Sýkora
RESULTS OF COMPARISON OF TWO RADON MONITORS 229 K.Holý, I.Bosá, T. Stanys, O.Holá, A.Polášková
ANALYSIS OF THE AVERAGE DAILY RADON VARIATIONS IN 233 THE SOIL AIR K.Holý, M.Matoš, R.Bôhm, T.Stanys, O.Holá, A.Polášková,
THE SURVEY OF DWELLINGS WITH INCREASED RADON 237 LEVELS IN SLOVAKIA M.Vičanová 10 21st RHD Jasná pod Chopkom
VARIATIONS OF RADON VOLUME ACTIVITIES IN SOIL AND 240 INDOOR AIR AND THEIR CORRELATION A.Mojzeš
THE POSSIBILITY OF EFFECTIVE AND EQUIVALENT DOSE 246 DETERMINATION IN THE NATIONAL PERSONNEL DOSIMETRY SERVICE J.Trousil, J.Plichta, J.Studená, J.Štrba
AUTHOR INDEX 247 2 ľ RHD Jasná pod Chopkom 11 SK98K0329
What are the purposes of Radiation Protection ? Milan Vladár
Institute of Preventive and Clinical Medicine Bratislava 12 21st RHD Jasná pod Chopkom
What are the purposes What is the Risk ? of Radiation Protection ? Health regulation to limit environmental Radiation protection is a risk, management and occupational exposures to radiation activity (Regulatory police). are based on the fact that it might cause Predictive model for the purposes of setting cancer, no matter how small regulatory policy must be based on Any exposure to radiation is harmful, and scientifical, economical, social and political one can calculate the probability of cancer considerations. from a linear extrapolation of observed Science is one of many inputs to risk cancer at high radiation exposures management.
Cancer risk and regulation Criticism of the LNT theory
The philosophy of cancer's risk has led to 1 ECONOMICAL: Regulatory compliance the widespread belief that there is no safe cost too much; negligible risk- reduction dose of radiation and that regulations benefits, should establish exposure limitation as SOCIAL: Public relations nightmare for low as possible if not zero the nuclear industry and other radiation Linear no-threshold (LNT) theory is used sources as a predictive model in regulatory SCIENTIFICALLY: Oversimplification of decision - making health risk in the low dose range
The scientific background for Minimum radiation dose and standard setting observational limitations • VWiat is the minimum radiation dose necessary to detect statistically At low doses of radiation (under 100 mGy) significant radiogenic cancer risk ? it is virtually impossible to detect a • Is there a dose threshold - can a single statistically increase in radiogenic cancer ionizing event in a critical cell result in These doses do not represent threshold cancer ? but reflect statistical limitations of • Is radiation damage repairable ? radioepidemiological studies. 21" RHD Jasná pod Chopkom 13
Lowest doses; of radiation The question of threshold (low LET) associated with cancer THRESHOLD SUPPORTERS: Health Population Lowest Reference 1 Effect Dose Threshold dose for carcinogenesis must exist (mGy) because DNA damage occurs at a high rate Total LSS 200 Sľifmlzu solid (1990) naturally cancer Leukemia LSS 200 Shlmlzu 1 (1990) Approximately 150 000 single-strand breaks Breast LSS, 200 Land and nitrogenous base lesions occur cancer Medically (1993) Irradiation spontaneously in every mammalian cell per Thyroid LSS, 100 Ron cancer Medically (1995) day Irradiation Dose 10 mGy/day adds only 20 events per day
The question of threshold Repair of radiation damage I. LNT SUPPORTERS: Single-strand breaks and base lesion occur • LNT theory predicts that radiation damage frequently and are readily repaired by the cell (risk) is cumulative; health effects should The spontaneous incidence of DNA double be independent of dose rate strand-break is - 0.01/cell; 10 mGy results in Radiobiological survey suggests that low a 40-fold increase frequency of these less dose-rate should be less effective than efficiently repaired lesions high dose-rate exposures There is linearity for DNA-dependent cell Several human epidemiological studies damage suggest that low dose rate reduces risk
Repair of radiation damage II. Beyond linearity 1 In radiation protection, predictions of health "Repair" has been convincingly demonstrated effects at occupational or environmental in cell culture systems and to lesser degree in doses (few mSv) are based on human epidemiological studies extrapolation from health effects observed "Repair" has been clearly documented at high at doses 200 mSv and higher doses (> 1 Gy); at exposure levels <10 mGy Two concepts are used by risk assessment: radiogenic effects are difficult to observe - Dose and Dose Rate Reduction Factors The importance of repair as a determinant of (DDREF) radiogenic cancer risk is still unknown - Negligible Individual Dose (NID) 14 2 ŕ RHD Jasná pod Chopkom
Dose and Dose Rate Dose -response analysis of Reduction Factors (DDREF) LSS leukemia mortality • ERR—U»~LQM— Smoottl 1» ERR = (a.D + P.D2) x S(D) 100 where ERR = (O/E)-1 / andS(D) = exp(-(c.D + d.D2)) *" 20 DDREF = /+ (pfa).D 12 3 4 Equivalent dost to tmrow JSv]
Dose -response analysis of Values of DDREF LSS solid cancer mortality ERR — LM — LOM — Smoot] I Parameter Leukemia Solid Cancers 2.8 a 0.87 0.47 J P . 4.50 0.0] Dose [Gy] DDRBF- ] +(p/a) x D 1 0.2 2.0 1.1 0.5 3.4 1.1 0 1 6.2 1.2 ( 1 i 4 1 Equvalent Negligible individual risk level Criticism of the DDREF (NCRP) The DDREF hypothesis cannot be supported by unrefuted finding of major low-dose • A negligible individual risk level (NIRL) is epidemiological studies: There are scan defined as a level of average annual human data that allow an estimate of the excess risk of fatal health effects given an DDREF (BEIR V.) annual value of 10~7, that correspond to an Studies of stable chromosomal aberrations annual effective dose of 0.01 mSv per following radiotherapy found that lower dose source or practice resulted in higher aberration yields per unit dose, exactly the opposite from what the • The NIRL is regarded as trivial compared DDREF would predict (aiduuuu.R<«.»i.i!i-as.i»i) to the health risk and can be dismissed from consideration 21" RHD Jasná pod Chopkom 15 Uncertainties in fatal cancer risk Operational limit estimation (NCRP No. 126) 1 Defining of 0.01 mSv as NIRL implies to PARAMETER COhTTRIBirnONTO some people that 0.1 mSv would not be MAGNITUDE VARIANCE Most negligible prohibit This is not that is intended i • ;.r,°/. -H'.""» Uiupcclkd uncemiitks 0.5- I.S 1.0 B.8V. 30.6% It might be quite useful for use a Trtntftr to US. popuktion 0.49. 2.35 1.00 6.6% 19.9% number such as 0.1 mSv or higher as Ufcline projection 0.62- 1.03 0.91 6.7% 0J% Statistical unccrtuntk* 0.5S- [.45 1.00 3.9% 4.2% an operational limit for requiring a Own! dosid etry onccrtstttkt 0.5<- I.M 0.Í4 J.ttt 4.3% practice to be examined MrciuiiQc(tJoa«rctitc«r 1.00- 1.19 1.10 B.i% 0.6% Probability distribution of the Conclusion I. risk coefficient per 1 Sv Parameter Population Workers ' No suggestion is made here that the Mean 3.99 E-2 3.69 E-2 nominal values of lifetime risk used for Median 3.38 E-2 3.16 E-2 radiation protection purposes should be changed (NRCP No. 126) Mode 2.36 E-2 2.22 E-2 • The LNT relationship, is at present the GSDEV 1.83 1.81 best tool to predict the risk probability of 90% confidenc 1.20 E-2 1.15 E-2 interval 8.84E-2 8.08 E-2 radiation at low doses (Beninson 1996) Conclusion II. Conclusion III. The evidence presented here suggest ' U.S. federal court dismissed all 2.100 that current standards for allowable lawsuits against GPU Nucl. Corporation radiation exposures - misunderstood by that claimed radiation injury from the the public as being "safe" - rather than 1979 TMI accident because: being unnecessarily restrictive, are in • the court determined that "at doses fact inadequate to safeguard public below 100 mSv, the causal link between health (Nussbaum 1998) radiation exposure and cancer induction is entirely speculative" 16 ŠK98K0330 21st RHDJasná pod Chopkom STATE SUPERVISION OVER RADIATION PROTECTION IN THE CZECH REPUBLIC Zdeněk Prouza State Office for Nuclear Safety Prague, Senovážné nám. 9, Czech Republic INTRODUCTION Presentation is aimed on: - the organisation of state supervision over radiation protection, - some aspects of the execution of state supervision over radiation protection domain. STRUCTURE AND COMPETENCIES OF THE REGULATORY AUTHORITY By the Act No. 85/1995 State Office for Nuclear Safety (SONS) is an integrated Regulatory Authority of the Czech Republic for nuclear safety and radiation protection with an independent budget and by Law No 18/1997 Cell. - Atomic Law - clearly declared competencies. SONS above all shall carry out: > state supervision over nuclear safety and radiation protection, the management of nuclear waste, spent fuel, nuclear materials, physical protection of nuclear facilities and nuclear materials (licensing, inspection systems) > regulation, evaluation of occupational, medical, public exposure due to practice or source within the practice, i.e. normal and potential exposure (determination of limits, constrains, guidance, clearance levels, etc.) > co-ordination of the State Radiation Monitoring Network and assurance of the international exchange of information on radiation situation, > professional co-operation with the International Atomic Energy Agency, y provide the Government and public with adequate information. From organisation point of view, three divisions were established in the framework of SONS, headed by Deputy-Directors and one independent department; our Division of Radiation Protection includes Departments of: • I Radiation Sources and Nuclear Power Plants (RSNPP), • Exposure Regulation (ER), • Management ofRad-Wastes and Radionuclides Releases (RWRR). Radiation hygiene departments of the Regional Hygienic Stations, Prague Hygiene Service and Uranium Industry's Institute of Hygiene at Přibram were transformed into the SONS Regional Centres, located in Prague, Plzeň, České Budějovice, Újstí nad Labem, Hradec Králové, Bmo and Ostrava, which are working in frame of RSNPP department. Radiation Hygiene Centre of the National Institute of Public Health in Prague was transformed into the SONS budgetary organisation - National Radiation Protection Institute (NRPI). LEGISLATION Due to political and economical changes in the Czech Republic whole legislation experiences an extensive reconstruction - the Act No. 18/1997 Cell. „Atomic Act" was 2ŕ RHD Jasná pod Chopkom 17 approved by the Government in December 1995, by Parliament of the Czech Republic in 24 January 1996 and came into force from IS' July 1997. In parallel with the „Atomic Act" twelve follow-up implementing Decrees were prepared by SONS; The Decrees on: a Requirements of Radiation Protection (including the problems of Radionuclides Release into the Environment, Radioactive Waste Management and Requirements for Reducing Population Exposures from Natural Sources), a Quality Assurance during Activities connected with Utilisation of Nuclear Energy, Transportation of Nuclear Materials and Radionuclide Sources, Q Type Approval of the Container Sets for Transports or Storage of Nuclear Materials and Radionuclide Sources, a Emergency Preparedness of Nuclear Facilities and Ionizing Radiation Sources in Case of a Radiation Accident, a Organization and Operation of the National Radiation Monitoring Network are particularly important for the radiation protection domain. The "Atomic Act" as well as these Decrees are based on the internationally adopted principles and recommendations of nuclear safety and radiation protection: • IAEAIBSS,No. 115/1994, • ICRP ReportNo. 60/1990, • EUDirective 96/29/EURATOM,etc. The "Atomic Act" and Decrees impose strong obligations upon users of sources, licensees; from standpoint of radiation protection: a) whoever performing radiation practices shall: • proceed in such manner that nuclear safety and radiation protection are ensured as a matter of priority. • ensure that his activities are justified by benefits outweighing risks from these activities, maintain a level of nuclear safety, radiation protection that the risk to life, health, environment shall be kept as low as reasonably achievable, • perform of intervention if the exposure can approach levels of acute damage to health, or if such measures are expected to provide more benefit than harm: • reduce exposure of people so that does not exceed the limits • have an implemented quality assurance system: b) licensees (for practice by "Atomic Act") above all shall: • assure radiation protection during all type of relevant practices with the sources for which they are authorized, • take immediate corrective actions if there are deviations from the approved licence's conditions, • submit to the SONS relevant (by "Atomic Act") documentation - monitoring and emergency plans, QA/QC programmes, etc., • prove to the inspectors that it fulfils all its stipulated duties in assuring radiation protection, • immediately inform the SONS and the corresponding inspector of serious facts, especially of emergency events which may affect radiation protection, • implement all measures imposed by inspectors or by the SONS headquarters. 18 2ľ'RHD Jasná pod Chopkom INSPECTION SYSTEM Three types of inspections and their evaluation are used: - regional inspections, which are planned and organised by Regional Centres at, by SONS headquarters determined, sources and practices, - inspections carried out by specialised on given inspection activities (nuclear facilities, uranium industry, nuclear medicine, users of accelerators) groups of inspectors; these inspections are planned and evaluated at SONS headquarters, ad hoc inspections (ad hoc inspection group is compiled from inspectors of SONS headquarters and Regional Centres) which are planned and evaluated at SONS headquarters and are concerned on from point of view radiation protection important facilities (NPP, rad-waste storage, large research centres, etc.). The execution of state supervision of radiation protection (ensured by 47 SONS inspectors for radiation protection) was mainly concentrated in source handling licensing procedures (evaluation of the level of compliance with the limits, conditions and requirements laid down by SONS licences - if the compliance was inadequate, the SÚJB specified requirements and conditions for continuation of given source application, practice) and in the inspections of workplaces where such activities are practised (agenda involved assessment of exposure to natural sources); during 1997 year: • 315 inspection visits were carried out at licensees in industry, • 1014 visits at licensees - users of sources for diagnostic and therapeutic purposes in human and veterinary medicine, • 604 at the other (research, education, services, etc.) users of sources- licensees. The basic findings of the SONS inspections on radiation protection during the last two years can be summarised as follows: - problems with implementation of new legislation - new types of documentation, implementation of QA/QC, ALARA programs, - despite of the privatisation process - the creation a number of small firms with limited personnel, technical and material possibilities - there was no global fall off in the radiation protection culture, - observance of the principles of radiation sources safe handling by some users has worsened; the area of concern is the sources distribution, since the situation in their records can be considered as somewhat confusing. During 1995-97 years (i.e. from re-organisation of Czech Radiation Protection) 56 events of the suspected loss of control over sources during transport or dismantling or as suspected theft or violent intrusion to workplaces maintaining such sources were reported. Although no health detriment was identified, radiation principles in some events had clearly been violated. The suspicion was confirmed in 27 events, these, however, were not very significant (the most serious event was the detection 1.5 TBq Go-60 source in a wagon transporting metal scrap - March 1996) from the radiation protection point of view; 29 events were evaluated as unsubstantiated reports, caused by various factors such as wrong measurements of exported metal scrap or the other transported material, measuring instrument malfunction, finding of insignificant radiation sources, and wrong recording. 2 ľ RHD Jasná pod Chopkom 19 LICENSING ACTIVITIES Licensing procedures for practice resulted in 1187 decisions by Regional Centres and 346 decisions issued by the SÚJB Headquarters during 1997 year. The licences mostly covered the handling, use of sources and the type approvals of sources. By the SONS the National register of ionising radiation sources up to now: > 8740 by SONS licensed X machines are registered, > 5376 equipment with sealed radionuclides source (106 in medicine; 50 medical radio-therapy units) are working under SONS licences, > 390 workplaces (47 departments of nuclear medicine) with unsealed radio-nuclides sources are licensed by SONS, ^ 7 power and research nuclear reactor units and 6 the other important facilities (with gamma irradiators for sterilisation or producing of radioisotopes) are working. Central Register of Occupational Exposure (CROE) registers 20,000 workers, which are controlled by the dosimetric services. Time trends of occupational exposure were analysed using data of the CROE; it can be concluded: • on medical applications of sources an advance of the number of workers was observed (in the radiodiagnostic it was employed 2000 more persons in 1997 than in 1990), • no significant changes of the effective dose (E) occurred, the values fit those ones observed in the developed countries; the average E are (in some occupational groups - radiodiagnostic) slight higher - innovations of technology, appearance of new diagnostic techniques, • on the field of industrial applications of sources (beyond the nuclear industry), the number of persons involved in non-destructive testing decreased (mainly due to privatisation of large companies); on the other hand, a new profession emerged, as the repair of SIR and maintenance; the trends in the average E exhibit no significant changes and also are at the same level as in industrially developed countries. Field Number of S E Emax Workers fmanSv] rmSvl [mSvl NPP 2433 1,28 0,53 20,4 Medicine 10525 9,58 0,91 120 Uranium industry 323 8,25 16,95 45,5 Others -3500 4,26 1,2 26,06 CONCLUSION The radiation protection in the Czech Republic has been from its beginning and it is up to now based on the same principles as in the other developed countries. It is possible to conclude that from the professional, technical as well as personnel standpoint, it is essentially provided at a relevant level. Due to changes in the economical and political spheres and in the organisational structure of state administration, the system of the Czech Republic Radiation Protection is now in phase of complete reorganisation: - new legislative system including ALARA, QA/QC programmes implementation should be introduced into daily practice of ionising sources users, - new, higher quality licensing and inspection system should be completely introduced and strengthened. 20 SK98K0331 21 RHD Jasná pod Chopkom RESULTS OF RADIATION PROTECTION AT NUCLEAR POWER PLANT DUKOVANY 1988 - 1997 Zdeněk Zelenka ČEZ, a. s. - Jaderná elektrárna Dukovany, 675 50 Dukovany Nuclear Power Plant (NPP) Dukovany has 4 units PWR WWER 440 MW each. The first unit was put in operation in May 1985 and the last one in July 1987. In July 1982 started to work the Laboratory of the Radiation Control of the Surrounding. In the structure of NPP Dukovany the Radiation Protection Department belongs to important factors ensuring reliable and safe operation of NPP Dukovany. The breaking down of the department corresponds to basic tasks of the department. They are as follows: * The Radiation Control of Operation ensures: > monitoring of technological systems, working environment and outlets by: - continual measuring dosimetry system - daily gamma spectrometric measuring of medium samples - continual measuring of ^N activity in the secondary circuit > check of keeping the general principles valid in the radiation protection * The Personal Dosimetry Control ensures: > monitoring of individual dose equivalents (IDE) and collective dose equivalents (CDE) of workers: • external exposure by: - film badge as the legal dosimeter - albedo dosimeter with ^LiF + ?LiF as the neutron dosimeter - thermoluminiscent dosimeter with alurninophosphate glass as the operational dosimeter - electronic dosimetry system as the secondary dosimetry system • internal contamination by: - whole body counter with HP Ge detector (gamma spectrometry) -13 li activity measuring in thyroid by NaI(Tl) detector - gamma spectrometric measuring of excretes - liquid scintilator (^H activity measuring in urine) > planning of exhausting CDE and IDE * The Radiation Control of Surroundings ensures: > balancing of outlets from NPP and watching the influence of them to surroundings by: - teledosimetry system around NPP in 20 km distance in 36 points 21" RHD Jasná pod Chopkom 21 - TLD system > low active samples measuring of environment, agricultural products food-stuffs by: - gamma spectrometry - proportional detector (90Sr activity samples measuring) - liquid scintilator (^H activity measuring in water samples) > co-operation in preparation of inhabitants for the case of accident From the start of NPP operation the trend has been to keep dose equivalents and outlets to surroundings on the lowest level as it is possible. Therefore the limit of the daily IDE was 2 mSv. The total number of cases with daily IDE>2 mSv is in the table 1. Table 1: Total Number of Cases with Daily IDE > 2 mSv Year 198 198 198 198 198 199 199 199 199 199 199 199 199 5 6 7 8 9 0 1 2 3 4 5 6 7 Number 0 0 3 3 0 0 0 0 2 1 0 3 0 The distribution of workers by the annual IDE interval for 1991 -1997 (Table 2) shows, that the low percent of workers is with IDE > 20 mSv but with IDE < 30 mSv. Table 2: Distribution of Number of Workers by Annual IDE interval Annual IDE interval [mSvl Year <0.1 0.1-0.5 0.5- 1-2 2-5 5-10 10- 15- 20- 25- 1 15 20 25 30 1991 375 168 29 12 27 2 0 0 0 0 1992 271 212 32 30 37 16 2 0 1 0 1993 592 337 86 77 34 21 6 0 1 1 1994 896 266 91 72 67 25 6 4 2 1 1995 945 288 105 82 73 31 13 10 1 0 1996 915 276 98 94 73 51 6 0 0 0 1997 800 300 116 100 100 32 3 4 1 0 In 1996 were accepted several precautions to bring down CDE of workers: - planning CDE on the month and on the each outage - direction of exhausting of IDE by hazardous works by the outages - daily CDE monitoring system - monthly CDE exhausting analysis. The low values of the CDE of workers (Table 3) and of the outlets (Table 4 and 5) show, that the technology is stabilised and the individual containments are functional. Only the ^H activity in the water samples of the river Jihlava is measured as the contribution to environment for NPP Dukovany (it is given by the reaction of neutrons with boron atoms in the primary circuit water). That fact is presented by the development of the inhabitant's CDE of outlets (Table 6). The model for inhabitant's CDE counting includes 1.023 mil. of population, living in 50 km distance from NPP 22 21st RHD Jasná pod Chopkom Dukovany. All presented data are taken from 1988, since all 4 units have been in operation. Table 3: Development of CDE and Number of Workers Year 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997 CED fSvl 1.31 1.60 1.02 1.27 1.87 1.78 1.41 1.69 1.45 1.52 N. of workers 2392 2017 2142 2262 2287 2574 2516 2489 2396 2339 Table 4: Development of Annual Gas Radioactive Outlets to Atmosphere - % Exhausting of Limit Year 1988 1989 1990 1991 1992 1993 1994 | 1995 1996 1997 RIG T % 1 0.032 0.015 0.002 0.006 0.027 0.103 0.106 0.130 0.077 0.010 Ra. iodines [ % 1 0.310 0.480 0.002 0.003 0.015 0.023 0.005 0.003 0.028 0.003 Aerosoles f% 1 0.020 0.103 0.055 0.056 0.118 0.116 0.082 0.103 0.047 0.136 Annual limits of gas outlets: Radioactive inert gases (RIG): 4.1 E15 Bq Radioactive iodines: 4.4 El 1 Bq Aerosols: 1.8EllBq Table 5: Development of Annual Liquid Radioactive Outlets to Water Flows - % Exhausting of Limit Year 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997 H-3[%1 59.5 85.9 91.4 83.2 87.7 84.5 70.9 65.8 78.8 66.9 A&FN f % 1 14.7 14.8 9.2 15.7 5.0 20.6 18.7 8.5 4.6 3.3 Annual limits of liquid outlets: Tritium (H-3): 2.2 El 3 Bq Activated a fissioned nuclides (A&FN): 2.0 E9 Bq Table 6:Development of Annual Values of Inhabitant's Exposure in NPP Surrounding of Outlets to Atmosphere and to Water Flows Year 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997 CDEofG.O. il.E-3Svl 0.3 0.3 0.1 2.4 1.7 1.7 1.5 2.8 1.9 1.0 CDEofL.O. ri.E-3Svl 36.4 44.0 40.8 40.5 44.1 51.5 38.4 37.9 41.9 30.0 Total CDE[l.E-3Svl 36.7 44.3 40.9 42.9 45.8 53.2 39.9 40.7 43.8 31.0 G. O. - gas outlets L. O.. - liquid outlets 2 ľ1 RHD Jasná pod Chopkom 23 The presented data of NPP operation shows, that NPP Dukovany reaches very good results in the Radiation Protection by comparison with NPPs from the advanced countries (the comparison CDE per 1 reactor unit or per 1 produced TWH - tables 7 and 8 [1]) and that its operation is safe and reliable. Table 7: Development of Average CDE per 1 reactor unit Average CDE per 1 Reactor [.Sv] Year 1988 1989 1990 1991 1992 1993 1994 1995 1996 All reactor types in the world 2.47 2.20 2.19 1.83 1.93 1.80 1.54 1.51 1.40 All PWR types in the world 2.45 2.29 2.23 1.92 1.92 1.72 1.37 1.49 L1.27 NPP Dukovany 0.33 0.40 0.25 0.32 0.47 0.45 0.35 0.42 0.36 Table 8: Development of Average CDE per 1 TWH Average CDE per 1 TWH [ Sv/TWH] Year 1988 1989 1990 1991 1992 1993 1994 1995 1996 All reactor types in the world 0.36 0.34 0.43 0.35 0.37 0.33 0.28 0.26 0.24 All PWR types in the world 0.45 0.42 0.39 0.33 0.32 0.28 0.22 0.23 0.20 NPP Dukovany 0.11 0.13 0.08 0.10 0.15 0.14 0.11 0.14 0.11 References: [1] - Occupational Exposures at Nuclear Power Plants 1986 -1996, ISOE, Sixth Annual Report, OECD-NEA, 1998 24 SK98K0332 21" RHD Jasná pod Chopkom ANALYSES OF OCCUPATIONAL DOSES AT THE DUKOVANY NPP Jiří Filip, Emil Fiala The State Office for Nuclear Safety, tř. kpt. Jaroše 5, 602 00 Brno Božena Jurochová, Zdeněk Zelenka Dukovany Nuclear Power Plant, 675 50 Dukovany The basic limits for workers and the guidance level for the annual coEective effective dose must not be exceeded - this is one basic presumption for a safe operation of nuclear facilities. The assumption for any reduction of occupational effective doses and for the implementation of the process of optimizing protection is the analysis of occupational exposure and its relation to all activities in a controlled zone of the NPP. In the Czech Republic, the guidance level for NPP's safe operation is characterized by a collective effective dose of 4 Sv for each GW of installed capacity [1], i.e., 4 x 440 MW at the Dukovany NPP. The radioactive waste depository and the interim spent fuel storage also contribute to collective effective dose at the Dukovany nuclear power plant site. To assess and analyse occupational effective doses of external exposure at the Dukovany NPP, the following parameters have been used [2]: - Annual collective effective dose S, - Average annual occupational effective dose EAVG (ALL) for all workers who enter the controlled zone, - Average annual occupational effective dose for all workers of the controlled zone who exceeded the recording level, EAVG^RL), - The percentage of workers NRE(>RL) exposed to the effective dose that exceeds 5mSv, - The percentage of collective effective dose SRE(>5 mSv) for workers that are exposed to the effective dose that exceeds 5 mSv, - The collective effective dose that is normalized to installed capacity (GW). The analysis of occupational effective doses is based on the results of personal dosimetry that are directly provided by the Dukovany NPP. The monitoring of occupational doses from external exposure is made by film dosimetry, TL and electronic dosimetry. TL and electronic dosimeters are evaluated operatively after planned operations; film badge dosimeters that are placed on reference points are used for the assessment of monthly occupational effective doses. To evaluate occupational effective doses, the value of dose equivalent in a point with a tissue depth of 10 mm - Hp(10) - is used. The annual recording level Hp(10) is a value of 0,1 mSv. The distribution of workers who enter the controlled zone versus annual occupational effective doses over the period of 1991 to 1997 and the other data which characterize annual effective doses are given in Table 1. The selected characteristics of the assessment are given in Table 2, e.g., average annual effective dose EAVG (>RL), coEective effective dose and the percentage of workers NRE exceeding the annual effective dose of 5 mSv and the fraction of collective effective dose caused by occupational effective doses exceeding 5 mSv and total coEective effective dose. The average values of annual occupational effective doses that exceed the recording level at 21" RHD Jasná pod Chopkom 25 workers in the controlled zone moved in a range from 1 to 1.86 mSv and collective effective doses were in a range from 1.28 to 1.85 Sv over the period of 1991 to 1997. The percentage of workers in the controlled zone at the Dukovany NPP versus annual occupational effective doses and the fraction of collective effective dose of the workers who exceeded the selected occupational effective dose versus total effective dose are shown in Table 3 and 4. From the results, you can see the contribution of occupational effective doses to the collective effective dose and the fraction of workers who contribute to higher effective doses. The annual effective dose of 5 mSv is exceeded by less than 2.6 percent workers and the three tenths of the basic limit for professionals (i.e., 15 mSv) by less than 0.5 percent. As in reference [3], we are convinced that 5 mSv is a more reasonable recording level than the three tenths of the basic limit for professionals. The fraction of occupational effective doses exceeding 5 mSv that contributes to the total collective dose is about 30 percent. From the overview of occupational effective doses in 1997, it is evident that the exceedance of a level of 5 mSv was only achieved at suppliers. The exceedance of 5 mSv level was achieved for 1.7 percent workers from suppliers, one worker even achieved a level of 20 mSv. The workers contributed by 23.8 percent to collective effective dose. In 1997, the contribution of the suppliers achieved 62 percent in the total number of workers who worked in the controlled zone, their contribution to collective effective dose is 74 percent. The resulting comparison of average occupational effective doses EAVG and collective effective doses S at the Dukovany NPP, which is given in table 5 and compared to a similar VVER type at the Jaslovské Bohunice NPP [3], is more favourable to the Dukovany NPP. The average values of collective effective doses that are normalized by a number of reactors and by installed capacity are. given in table 6. The average collective effective dose related to one GW of installed capacity is about 25 percent of the guidance level for the NPP's safe operation, see the regulation [1]. The analysis of occupational doses is the important basis for the ALARA principle [4]. The results of personal dosimetry and the analysis of all activities in the controlled zone were used at the Dukovany NPP over two last years to schedule monthly collective effective doses. For the selected activities where the exceedance of a level of 1 mSv is possible in occupational effective doses on operative dosimeters, the analyses of radiation protection provisions are needed. Because suppliers mainly contribute to the collective effective dose, our attention should focus on these workers who should be involved in the reduction of their contribution to the above-mentioned doses. REFERENCES: [1] Regulation of the State Office for Nuclear Safety, no. 184/1997 Sb. on Requirements for Radiation Protection Provisions, „A Digest of Laws of the Czech Republic", part 66 (1997), pp. 3962-4095. [2] UNSCEAR 1992, Occupational Radiation Exposures, United Nations Sale, NY (1992). [3] VLADÁR, M. - NIKODÉMOVA, D. - HUTTA, J. - MOCKO, S.: Analyses of occupational dose distributions at the Jaslovské Bohunice NPP, ,3ezpe5nost jaderné energie", 5 (43), 1997, pp. 239-242 [4] Basic Safety Standards for Protection against Ionizing Radiation, Safety Series No. 115-1, IAEA Vienna (1994) 26 2 ľ' RHD Jasná pod Chopkom Table L The number of workers, their annual effective dose E, collective effective dose S , and average occupational effective doses EAVG at the Dukovany NPP over 1991to 1997 E fmSv] 1991 1992 1993 1994 1995 1996 1997 1991-1997 <0.1 762 664 1,121 1,659 1,579 1,507 1,345 8,637 0.10 - 0.49 860 880 934 454 422 426 494 4,470 0.50-0.99 182 180 184 140 169 145 179 1,179 1.00 -1.99 134 137 160 125 150 154 146 1,006 2.00-4.99 117 168 108 94 111 105 135 838 5.00 - 9.99 10 48 55 30 34 53 32 262 10.00 -14.99 0 4 9 6 13 6 3 41 15.00 -19.99 0 0 1 5 10 0 4 20 ž 20.00 0 1 2 3 1 0 1 8 ALL 2,065 2,082 2,574 2,516 2,489 2,396 2,339 16,461 N(>RL) 1,303 1,418 1,453 857 910 889 994 7,824 S[Sv] 1.28 1.85 1.78 1.41 1.69 1.45 1.52 10.98 EAVG(ALL) fmSvl 0.62 0.89 0.69 0.56 0.68 0.61 0.65 0.67 EAVG(>RL) fmSv] 0.98 1.30 1.22 1.65 1.86 1.63 1.53 1.40 RL - Recording Level Table 2. Characteristics of occupational external exposures determined from the annual data of personal dosimetry Parameters 1991 1992 1993 1994 1995 1996 1997 1991-1997 S[Sv] 1.28 1.85 1.78 1.41 1.69 1.45 1.52 10.98 EAVG(>RL) fmSvl 0.98 1.30 1.22 1.65 1.86 1.63 1.53 1.4 NRE(E:>5 mSv) \%\ 0.5 2.5 2.6 1.7 2.3 2.5 1.7 2 * SRE(E;>5 mSv) [%] 4.88 21.3 31.5 30.6 35.7 29.8 23.8 Table 3. The percentage of the Dukovany NPP's workers with NRE(žE) who exceeded the annual occupational effective dose E Annual occupationaleffective dose E [mSv] Year aO.l >0.S ;>1 ;>2 >5 £10 žl5 ;>20 2:50 1991 63.1 21.5 12.6 6.2 0.5 0.0 0.0 0.0 0.0 1992 68.1 25.8 17.2 10.6 2.5 0.2 0.0 0.0 0.0 1993 56.4 20.2 13.0 6.8 2.6 0.5 0.1 0.1 0.0 1994 34.1 16.0 10.5 5.5 1.7 0.6 0.3 0.1 0.0 1995 36.6 19.6 12.8 6.8 2.3 1.0 0.4 0.0 0.0 1996 37.1 19.3 13.3 6.8 2.5 0.3 0.0 0.0 0.0 1997 42.5 21.4 13.7 7.5 1.7 0.3 0.2 0.0 0.0 2 ľ' RHD Jasná pod Chopkom 27 Table 4. The percentage of the Dukovany NPP's workers with collective effective dose SRE(>E) that exceeds the annual occupational effective dose E Annual occupational effective dose E [mSv] Year >0.1 >0.S >1 ä2 >5 žlO žl5 >20 >50 1991 98.4 76.8 54.1 35.7 4.88 0.0 0.0 0.0 0.0 1992 99.2 86.3 67.5 54.7 21.3 3.8 1.3 1.3 0.0 1993 98.6 78.9 65.8 51.3 31.5 10.1 2.9 2.9 0.0 1994 97.4 77.7 67.6 52.7 30.6 16.3 9.7 5.1 0.0 1995 98.1 81.3 71.3 57.0 35.7 20.7 11.7 1.4 0.0 1996 97.8 80.4 70.1 53.8 29.8 4.75 0.0 0.0 0.0 1997 98.3 79.7 68.8 52.8 23.8 8.65 6.4 2.7 0.0 Table 5. Comparison of average annual effective doses EAVG and collective effective doses S for the Dukovany NPP's workers (see table 1) and the Jaslovské Bohunice NPP's workers [4] 1990 1991 1992 1993 1994 1995 1996 1997 Parameters Dukovany NPP SfSvl * 1.28 1.85 1.78 1.41 1.69 1.45 1.52 EAVG(>RL) fmSvl * 0.98 1.3 1.22 1.65 1.86 1.63 1.53 Jaslovské Bohunice NPP [4] S[Svl 1.83 3.14 4.53 4.18 2.19 * * * * * * EAVG(>RL) [mSvl 1.72 2.40 2.16 2.09 1.51 Table 6. Collective effective dose (CED) normalized by the number of reactors and installed capacity Installed CED CED Number of Year capacity per reactor per installed reactors [GW1 [Sv] capacity [Sv/GW] 1991 4 1.76 0.32 0.73 1992 4 1.76 0.46 1.05 1993 4 1.76 0.45 1.01 1994 4 1.76 0.35 0.80 1995 4 1.76 0.42 0.96 1996 4 1.76 0.36 0.82 1997 4 1.76 0.38 0.86 The value for safe operation in the Czech Republic [1]: 4.00 28 SK98K0333 2ŕ RHD Jasná pod Chopkom MONITOROVANIE AKTIVITY OSÔB A VOZIDIEL NA VÝSTUPE Z SE-EBO JASLOVSKÉ BOHUNICE Ľubomír Dobiš, Ján Kaizer, Jaroslav Svitek Slovenské elektrárne, a.s. Atómové elektrárne Bohunice, o.z. 919 31 Jaslovské Bohunice Obsah: 1. Úvod. 2. Technický popis a umiestnenie monitorov osôb. 3. Technický popis a umiestnenie monitorov vozidiel. 4. Výsledky monitorovania počas prvých šiestich mesiacov. 5. Diskusia výsledkov. 6. Záver, 1. Úvod V januári 1998 bol v SE-EBO do prevádzky uvedený systém monitorovania aktivity osôb a vozidiel na výstupe z JE. Monitorovací systém plní dva základné ciele: a, primárny - v súčinnosti s personálom ochrany podniku slúži na kontrolu a zabránenie neautorizovanému vyneseniu rádioaktívnych žiaričov z JE b, sekundárny (tzv. ochrana do hĺbky) - vytvára redundanciu merania kontaminácie na výstupe zkontrolovaného pásma pri prípadnom zlyhaní (poruche) monitorov na hygienických slučkách alebo pri zlyhaní ľudského faktora (porušenie predpisov radiačnej bezpečnosti) Na JE VI projekt vôbec neuvažoval s monitorovaním osôb resp. vozidiel na výstupe z JE. Uvedený nedostatok riešil odborný útvar zabudovaním okienkových GM trubíc do rámov vstupných dvier. Prevádzka JE V2 už disponovala labyrintovým výstupom osôb s rámom osadeným GM trubicami a pre monitorovanie vozidiel jedným scintilačným detektorom. Uvedené zariadenie nemohlo spĺňať sekundárny cieľ monitorovania, pretože ich RDA (Reliably Detectable Activity) bola najmä z dôvodu obmedzenej detekčnej doby podstatne vyššia ako monitorovacieho zariadenie na výstupe z kontrolovaného pásma. V súčasnosti monitory osôb a vozidiel doplnené o systém organizačných opatrení tvoria súčasť systému zabezpečenia kvality SE -EBO. 2. Technický popis a umiestnenie monitorov osôb Monitor PM7 na kontrolu osôb opúšťajúcich areál SE-EBO dodala ry Canberra Badín od výrobcu Eberline USA. Technický popis Monitor, ktorý bol upravený v rámci predkomplexného vyskúšania obsahuje: • detektory: • 6 veľkoobjemových plastikových scintilátorov (hlava, nohy, 2x2 bočné) • detektory sú z vonkajšej strany tienené Pb • elektroniku a signalizačný zvukový a svetelný panel redukovaný na dve farebné polia: • červený - prekročenie nastavenej úrovne aktivity 2ľ RHD Jasná pod Chopkom 29 • zelený - meranie Ostatné signalizačné prvky prístroja sú vypnuté resp. nefunkčné. Na paneli je konektor pre pripojenie k PC. • software: kompletný softwarový balík pre kontrolu, nastavenie a kalibráciu prístroja, tlač kalibračného protokolu, prenos po RS232 Nastavená a kontrolovaná RDA 9,25 - 10,4 kBq Csl37 pri dobe merania ls. V rámci úprav počas skúšok bola snaha minimalizovať dobu merania detektora pri zachovaní čo najväčšej citlivosti detektorov. Umiestenie monitorov Správne umiestnenie monitorov je dôležité hlavne s pohľadu efektívneho zásahu pracovníkov ochrany pri signalizácii monitora. Zároveň je nutné podotknúť, že projekt umiestnenia monitorov zasahoval do existujúceho a funkčného systému AKOBOJE (systém ochrany podniku), ktorý nemohol byť podstatným spôsobom narušený. Dôležitou požiadavkou bolo nenarušiť plynulosť odchodu pracovníkov po skončení pracovnej smeny. Umiestnenie monitorov na vrátnici JE VI a V2 znázorňuje uvedená schéma: Pracovník ochrany podniku |Turnikety| Monitor PM7 Priestor osobnej vrátnice \ 3. Technický popis a umiestnenie monitorov vozidiel Monitor FHT 1341 určený na kontrolu nákladných vozidiel opúšťajúcich areál SE- EBO.VYZ dodala fy Caberra Backn od výrobcu Eberline USA. Technický popis Monitor obsahuje: • dva veľkoplošné detektory (1000x500x50 mm) umiestnené po stranách vozovky • riadiacu a meraciu jednotku FHT 1100 • systém senzorov na detekciu meraných vozidiel • signalizačnú jednotku • software SCRAP Nastavená a kontrolovaná RDA je nižšia ako 550 kBq v strede vozovky (pri pohybe vozidiel cca 15 km/hod). Režim merania vozidla spúšťajú a odstavujú fotoelektrické senzory umiestnené na stojane detektora. 30 2 ľ1 RHD Jasná pod Chopkom Umiestnenie monitorov Detektory monitorov boli postavené po stranách vstupno-výstupnej vozovky vedľa vrátnice JE V2 a vedľa vrátnice JE Al tak, aby kraj detektora lícoval s krajom vozovky. Toto umiestnenie nie je ideálne z pohľadu detekčnej citlivosti t.j. automobil nie je v geometrickom strede medzi detektormi počas kontroly, ale opätovne sa projekt musel prispôsobiť stávajúcemu stavu ako v prípade monitorov osôb. Uvedený problém vyrieši vybudovanie novej nákladnej vrátnice, ktorá nahradí meranie na JE Al. Signalizačný panel a jednotka FHT 1100 sú umiestnené priamo na stanovišti pracovníkov ochrany podniku. 4. Výsledky monitorovania počas prvých šiestich mesiacov Výsledky monitorovania sú zhrnuté do tabuľky alarmných signalizácií: Prehľad alarmných signalizácií na monitoroch PM7 - monitorovanie osôb Kvaiifilriwanr. PoniSpnie nravirlip 1 raHiařnpi Annmáli Ternni Elektráreň Kontaminácia Kontaminácia pracovníka (vnút. predmetov kontaminácia,telo,šaty) VI 16 2 L i 1 V2 1 1 l - Al 3 - 1 2 1 Prehľad alarmných signalizácií na monitoroch FHT 1431 - monitorovanie vozidiel Kvalifikované Pnrníenie nraviriinl rad. Annmália Elektráreň Kontaminácia vyvážaného Stav. materiál, materiálu suroviny V2 1 - Al 1 14 Pričom pod výrazom - • kontaminácia pracovníka rozumieme, že následné meranie potvrdilo kontamináciu rádioaktívnymi látkami na tele, šatoch prípadne vnútornú kontamináciu, najčastejšie však ich kombináciu • kontaminácia predmetov - kontaminácia bola potvrdená na nesených predmetoch • anomália - kontaminácia nekvalifikovateľná pracovnými predpismi SE-EBO (napr. ciferník hodiniek, aktivita KOH, kontaminácia stavebných látok prírodnými izotopmi pri dovoze stavebného materiálu do SE-EBO) • terapia - vnútorná kontaminácia spôsobená lekárskou rádionuklidovou terapiou • kontaminácia vyvážaného materiálu - kontaminácia materiálu použitého v prevádzke JE a 21 RHD Jasná pod Chopkom 31 Z uvedeného je zrejmé, že zvýraznené pojmy museli byť kvalifikované ako porušenia pravidiel radiačnej bezpečnosti. 5. Diskusia výsledkov Poznamenávame, že cieľom analýzy získaných výsledkov nie je určenie príčiny (zdroja) kontaminácie, ale popis redundantného systému monitorovania na vrátniciach areálu JE Bohunice. Šetrenie príčiny a následkov kontaminácie a návrhy nápravných opatrení sú v kompetencii odborných útvarov SE-EBO a nie sú predmetom tohto článku. Z pohľadu na tabuľky vystupujú do popredia najmä nasledovné skutočnosti: a, potvrdil sa očakávaný predpoklad, že monitory povrchovej kontaminácie na hygienickej slučke už nevyhovujú súčasným kritériám na meracie prístroje daného zamerania a to ako z pohľadu citlivosti tak z pohľadu rozsahu kontroly povrchu tela pracovníkov (relatívne malá plocha okienka použitých GM trubíc, nedostatočné monitorovania hlavy a paží pracovníkov - nevyhovujúca geometria merania) b, z pohľadu na časový priebeh signalizácií monitorov PM7 na vrátnici JE V-l vidieť, že zavedenie monitorov zvýšilo disciplínu merania a prechodu cez HS - po relatívne veľkom počte signalizácií po zavedení monitorovania postupne počet signalizácií poklesol. Opätovné zvýšenie signalizácie v závere hodnoteného obdobia môžeme pripísať začiatku GO 2. bloku a s tým súvisiacemu príchodu veľkému množstvu neskúsených pracovníkov cudzích firiem (viď nižšie uvedený graf). Časový priebeh signalizácií monitora PM7 na JE VI. 4 • # •S 2 • 1 • —i- \ i 1 1 1 h"H A 1 1 1 i 1 h"+ i 1 1 1 0- « vi .-• vi »' <*i *ó © d r* ~- ví o* ~ n *- n D 6. Záver Na základe skúseností z polročnej prevádzky monitorov rádioaktivity na výstupe z areálu SE-EBO možno konštatovať, že uvedená investícia preukázala svoj zmysel. Predkladané výsledky jedhoznačne poukazujú na veľký význam rekonštrukcie a modernizácie JE aj 32 21st RHD Jasná pod Chopkom v oblasti radiačnej bezpečnosti. Výsledky monitorovania na výstupe z JE dodali silné argumenty aj na urýchlenie modernizácie monitorovania povrchovej kontaminácie na hygienickej slučke - výmena zastaralých monitorov na hygienických slučkách JE VI a V2 - ktorá v súčasnosti prebieha. Nové monitory kontaminácie na HS budú spĺňať všetky požiadavky súčasného poznania v oblasti radiačnej bezpečnosti a ich inštaláciou stúpne kvalita monitorovania na výstupe z kontrolovaných pásiem tak, že frekvencia záchytu kontaminácie na výstupe z SE-EBO klesne na nulu. 2ľ'RHD Jasná pod Chopkom SK98K0334 33 QA REQUIREMENTS FOR CZECH NUCLEAR ACTIVITIES AND RADIATION PRACTICES Jarchovský Daniel State Office for Nuclear Safety, Czech Republic Abstract In the QA area, the State Office for Nuclear Safety ensures: • preparation of legal regulations • reviewing and evaluation of the licensee's safety related documentation and Quality Assurance Programmes • inspections to establish compliance with the legal regulations • provides interpretation of these legal regulations, with subsequent elaboration of specific requirements into the form of recommendations and guidelines LEGISLATION FORMER LEGISLATION Act No 28/1984 Coll. on State Supervision over Nuclear Safety of Nuclear Installations. Legislative Rules CSKAE Decree No 5/1979 Classified Items Contents of the Quality Assurance Programme (QAP) Procedure for the QAP creation and approval Control activities Principles of the quality assurance Designing Manufacturing, supply, assembling and start-up Operation and repairs CSKAE Regulation No 436/90 Coll. • Classified Items and their ranking • Principles of the quality assurance • Quality Assurance documentation • Procedure for the QAP creation and approval • Requirements for quality assurance in the individual stages CURRENT LEGISLATION Act Law No 18/1997 Coll. on peaceful utilisation of nuclear energy and ionising radiation (the Atomic Act) including amendments and supplements of related acts 34 21" RHD Jasná pod Chopkom Basic provisions (§ 4, para 7): " Any person, performing or providing for practices related to nuclear energy utilisation or radiation practices, must have an implemented quality assurance system, in the extent and in a manner set out in implementing regulation, aiming at achieving the required quality of a relevant item, including tangible and intangible products, processes or organisational arrangements, with respect to the importance of this item from the aspect of nuclear safety and radiation protection." Other provisions § 13, para 5 - An approval issued by the Office of a quality assurance programme for the licensed practice is a prerequisite for issue of a licence.... § 17, para 1, letter d) - A licensee shall, besides other obligations established in law comply with the approved quality assurance programme... Legislative rules SUJB Regulation No 214/1997 Coll. Key points of the new Regulation contents: a) Criteria for the classified items and their ranking into safety classes, b) Requirements for the extent and procedures of the Quality System implementation for practices connected with utilisation of nuclear energy and radiation practices, c) Requirements for quality assurance of classified items with regard to their ranking into safety classes, d) Requirements for the contents of Quality Assurance Programmes and the List of classified items, e) Requirements for the quality system of licensed practices NEW REGULATION IS BASED ON THE FOLLOWING DOCUMENTS: • IAEA Safety Series No. 50-C/SG-Q : 1996, • ČSN ISO 9001: 1994, • ČSN ISO 9004-1 :1995, • Former ČSKAE Decree No. 436/90 Coll. Main difference between the former and new Regulation is a much wider scope of the new one: • wider range of practices for which the requirement that the practices shall be performed in compliance with the implemented and documented quality assurance system, applies, • complex requirements for the quality system with regard to processes important to nuclear safety and radiation protection (not only to classified items). 2ľ RHD Jasná pod Chopkom 35 CONTENTS OF REGULATION No. 214/1997: 1. Introductory provisions Subject and extent Implementation of a quality assurance system . Requirements to a quality assurance system Basic requirements Controlled conditions of the processes Procurement, contracts Design Documentation, data Identification of the products Inspection and test procedures Calibration and handling of the control and measuring devices Records Checks (inspections) Personnel Design of a nuclear installation Important processes and areas of the nuclear installation operation Specific processes Evaluation (internal and independent audits) 3. Requirements for the QA of classified items 4. Requirements for the contents of QAP 5. Criteria for ranking of classified items into the safety classes 6. Extent and the format of the classified items list 7. Temporary and concluding provisions CONCLUSION The SUJB strategy in its regulatory activity in the QA area, reflected in the new legislation, is based on a gradual transition from a formal evaluation of quality assurance programmes to supervising activities, conform to QA performance approach which means, besides other, that an appropriate responsibility for the quality of work is imposed on those persons who actually perform and supervise the work. 36 2 ŕ RHD Jasná pod Chopkom SK98K0335 INFORMACE O SPRÁVĚ ÚLOŽIŠŤ RADIOAKTIVNÍCH ODPADŮ Věra Starostová Správa úložišť radioaktivních odpadů, Charvátova 6, 110 00 Praha 1 Správa úložišť radioaktivních odpadů - SURAO —jako státní organizace zřízená zákonem č. 18/1997 Sb. (atomovým zákonem) za účelem zajišťování činností spojených s ukládáním radioaktivních odpadů má za sebou více než rok své existence. V příspěvku je diskutován jak předmět činnosti SÚRAO, tak i jeho postupné naplňování. Dále jsou uvedeny ty povinnosti jednak žadatelů o povolení Státního úřadu k některé z činností podle § 9 atomového zákona, jednak držitelů těchto povolení, jejichž naplňování se váže také k činnosti SURAO, a je podána informace o nejvhodnějším způsobu komunikace se SÚRAO. 21" RHD Jasná pod Chopkom SK98K0336 37 WHAT IS MORE DANGEROUS: NUCLEAR POWER PLANTS OR CARBON FIRED POWER PLANTS? Jozef Kuruc Department of Nuclear Chemistry, Faculty of Natural Sciences, Comenius University, 84215 Bratislava, Slovak Republic The aim of this paper is to compare environmental impacts of radionuclides and other pollutants released into environment from nuclear power plants (NPP) and coal fired power plants (CFPP). Global energy in 1993 reached 338 exajoules; it is 40% greater than in 1973 [1]. In the next table primary energy consumption by fuel in 1997 is compared [2]. Table 1. Primary energy consumption by fuel in 1997 (million tons of oil equivalent) Oil Natural Gas Coal Nuclear Hydroelectri Total Energy c power 3395.5 1977.3 2293.4 617.4 225.4 8509.2 39.90 % 23.24 % 26.95 % 7.26 % 2.65 % Coal combustion continues to be dominant fuel source for electricity production. Fossil fuel's share has decreased from 76.5% in 1970 to 66.3% in 1970, while nuclear energy's share in the global electricity has increased from 1.6% in 1970 to 17.4% in 1990. Today, NPPs provides about 18% of the world's electricity 340.347 MWe [3]. It offers an important environmental advantage in that it produces no harmful emissions as carbon dioxide (CO2), sulphur dioxide (SO2) or nitrogen dioxide (NOX) in normal operation. According to calculation, if the electricity currently generated by nuclear power plants globally were, instead, to be produced by burning coal, another 8%, or 1.6x10 tons of carbon dioxide, would be injected into the earth's atmosphere (annually). World-wide, there were 443 nuclear power reactors in operation at the start of 1997, producing about as much electricity as obtained from hydropower [3]. A large power reactor (1000 MWe) uses 150 tons of natural uranium a year, equivalent to 2.5 million tons of black coal or 12 million barrels of oil (1.908 million m3 of oil) [4]. The supply of nuclear power is to increase 2.5 times between 1989 and 2010 with a target of 30% of all electricity being generated by nuclear power by the year 2010 [4]. At the combustion of coal and other fossil fuels arising enormous amounts of greenhouse gases, as carbon dioxide, nitrogen oxides and other pollutants, as sulphur dioxide. The increasing use of nuclear power since 1960s, combined with steady increases hydropower, have helped curb world-wide carbon dioxide output. If the electricity energy generated worldwide each year were produced in NPPs instead by coal-pówered plants, there would be additional emissions of 1,600 million tons of CO2. If the world were not employing nuclear power today, global carbon dioxide emissions would be at least 8% greater every year. 38 2ŕRHD Jasná podChopkom Nuclear power is exceptionally clean in operation. Concern is usually focused on the highly toxic and radioactive spent fuel and nuclear wastes. In addition to their toxicity and radioactivity, they are limited in volume, which facilitates waste disposal. This contrasts sharply with the waste disposal problem for fossil-fuelled plants. For example: A 1000 MWe CFPP with optimal pollution abatement equipment will emit into the atmosphere 900 tons of SO2 per year; 4500 tons of NOX; 1300 tons of particulates; and 6.5 million tons of CO2. Depending on the quality of the coal, up to one million tons of ashes containing hundreds of tons of toxic heavy metals (arsenic, cadmium, lead) and natural radionuclides (uranium, thorium, daughter radionuclides, K-40) will have to be disposed of [5]. By contrast, aNPPof 1000 MWe capacity produces annually some 35 tons of highly radioactive spent fuel. If the spent fuel is reprocessed, the volume of highly radioactive waste will be about 3 m\ The entire nuclear chain supporting this 1000 MWe plant, from mining through operation, will generate, in addition, some 200 m3 of intermediate level waste and some 500 m3 of low level waste per year [5]. In 1992, global emissions of CO2 - the prime greenhouse gas added to the atmosphere as a direct result of human activity - amounted to 26.4 x 109 tons per year, of which 84% (22.3 x 109 tons) was from industrial activity. Emissions from industrial activity have climbed 38 percent over the past 20 years. The United States had the highest per capita emissions -19.1 tons per year - among the nations that were the major sources of global emissions in 1992 [1]. In the Slovak Republic the combustion of fossil fuels is a source of about 94 % emissions of carbon dioxide, ~58.3 million tons CO2 annually. Slovakia with its production of 11 tons of CO2 per capita in 1990 highly had surpassed then European mean value 7.3-tons/capita.year [6]. With regard to environmental impacts, nuclear power offers specific benefits. In routine operation, nuclear power plants and the fuel cycle facilities do release small quantities of radionuclides, mostly there are radioisotopes of inert gases Xe, Kr and iodine. However, the rules developed and implemented several decades ago for limiting radioactive emissions satisfy criteria for protecting human health and more than adequate to protect the environment. The other emissions, residuals, and burdens from NPPs and fuel cycle facilities are lower than those arising from fossil-fuel electricity generations chains and comparable or lower than those from renewable energy systems. Taking into account the entire up-stream and down-stream energy chains for electricity generation, nuclear power emits 40 to 100 times less CO2 than currently used fossil-fuel chains [10]. Greenhouse gas emissions from the nuclear chain are due mainly to the use of fossil fuels in the extraction, and enrichment of uranium and to fuels used in the production of steel, cement and other materials for the construction of reactors and fuel cycle facilities. These emissions, which are negligible relative to those from the direct use of fossil fuels for electricity generation, can be reduced even further by energy efficiency improvement [12]. From a point of view conservation of the environment it is important what area would be need for a power plant. This value depends from energy density: 1 kg of firewood produces about 1 kWh of electricity; 1 kg of coal produces about 3 kWh of electricity; 1 kg of oil produces about 4 kWh of electricity; 1 kg of natural uranium produces about 50 MWh of electricity; and 1 kg of plutonium produces about 6 GWh. The low energy density of the renewable sources means that if we want significant amounts of energy (electricity) from them, we must "harvest" them over 21" RHD Jasná pod Chopkom 39 large areas - and this is very expensive. It has been calculated that to achieve the electricity generating capacity of a 1000 MWt power plant, we would need: an area of 50 to 60 km2 to install solar cells or windmills, or an area of 3000 to 5000 km2 to grow the needed biomass [5]. By contrast, we would need an area of only a few square kilometres for a NPP, including all of its fuel cycle requirements. Coal is one of the most impure of fuels. Its impurities range from trace quantities of many metals, including natural radionuclides uranium and thorium, to much larger quantities of Al and Fe to still larger quantities of impurities such as sulphur. Products of coal combustion include the oxides of carbon, nitrogen, and sulphur; carcinogenic and mutagenic substances; and recoverable minerals of commercial value, including nuclear fuels naturally occurring in coal. Coal ash is composed primarily of oxides of silicon, aluminium, iron, calcium, magnesium, titanium, sodium, potassium, arsenic, mercury, other metals, and sulphur plus small quantities of uranium and thorium. Fly ash is primarily composed of non- combustible silicon compounds (glass) melted during combustion. Tiny glass spheres form the bulk of the fly ash. Since the 1960s particulate precipitators have been used by CFPPs to retain significant amounts of fly ash rather than letting it escape to the atmosphere. When functioning properly, these precipitators are approximately 99.5% efficient. Utilities also collect furnace ash, cinders, and slag, which are kept in cinder piles or deposited in ash ponds on coal-plant sites along with the captured fly ash. Trace quantities of uranium in coal range from less than 1 ppm in some samples to around 10 ppm in others. In some cases a concentration of uranium can to amount up to 1000 ppm [11] and they can be used as uranium resources. Generally, the amount of thorium contained in coal is about 2.5 times greater than the amount of uranium. For a large number of coal samples, according to Environmental Protection Agency figures released in 1984, average values of uranium and thorium content have been determined to be 1.3 ppm and 3.2 ppm, respectively. Using these values along with reported consumption and projected consumption of coal by utilities provides a means of calculating the amounts of potentially recoverable breedable and fissionable elements. The concentration of fissionable U-235 (the current fuel for NPPs) has been established to be 0.71% of U content. In the coals mined in the Slovakia (3.8 million tons in 1996) concentrations of natural radionuclides were determined in previous years, but these values so far were not published [7]. And what about —3.8 million tons of coal imported into Slovakia yearly? Assume that the typical plant has an electrical output of 1000 MW. Existing coal- fired plants of this capacity annually burn about 4 million tons of coal each year. Using these data, the releases of radioactive materials per typical plant can be calculated for any year. For the year 1990, assuming coal contains uranium and thorium concentrations of 1.3 ppm and 3.2 ppm, respectively, each typical plant 1000 MWe released 5.2 tons of U (containing 36.92 kg of U-235) and 12.8 tons of Th that year. Total releases in 1990 from worldwide combustion of -3300 million tons of coal totalled -4552 tons of uranium (containing -32317 kg of U-235) and -10860 tons of thorium. These values continually increased! Based on the predicted combustion of 12,580 million tons worldly during the year 2040, cumulative releases for the 100 years of coal combustion following 1937 are predicted [9] to be Planetary release (from combustion of 637,409 million tons): Uranium: 828,632 tons (containing 5883 tons of U-235); Thorium: 2,039,709 tons. 40 2 ľ' RHD Jasná pod Chopkom On the base these amounts McBridge et al. concluded that „Americans living near CFPPs are exposed to higher radiation doses than those living near NPPs that meet government regulation [9]. The main sources of radiation released from coal combustion include not only uranium and thorium but also daughter products produced by the decay of these nuclides, such as isotopes of radium, radon, polonium, bismuth, and lead. Although not a decay product, naturally occurring radioactive potassium-40 is also a significant contributor. According to the National Council on Radiation Protection and Measurements (NCRP), the average radioactivity is 427 uCi/t of coal. This value can be used to calculate the average expected radioactivity release from coal combustion. For 1990 the total release of radioactivity from worldwide 3300 million tons coal combustion was, therefore, about 1.41 MCi. For comparison, according to NCRP Reports No. 92 and No. 95, population exposure from operation of 1000-MWe NPP and CFPPs amounts to 4.90 man-Sv/year for CFPPs and 0.048 man-Sv/year for NPPs. Thus, the population effective dose equivalent from CFPPs is 100 times that from NPPs. For the complete nuclear fuel cycle, from mining to reactor operation to waste disposal, the radiation dose is cited as 1.36 man-Sv/year [13,14]; the equivalent dose for coal use, from mining to power plant operation to waste disposal, it is probably unknown. During combustion, the volume of coal is reduced by over 85%, which increases the concentration of the metals originally in the coal. A global average concentration of uranium is 10 ppm in ash [15]. Although precipitators retain significant quantities of ash, heavy metals such as uranium and thorium tend to concentrate on the tiny glass spheres that make up the bulk of fly ash. This uranium is released to the atmosphere with the escaping fly ash, at about 1.0% of the original amount, according to NCRP data. The retained ash is enriched in uranium several times over the original uranium concentration in the coal because the uranium, and thorium, content is not decreased as the volume of coal is reduced. Fly ash (with U and Th and daughter radionuclides are precipitated with snow and rain and increased activity of air and Earth's surface. Another unrecognised problem is the gradual production of Pu-239 through the exposure of U-238 in coal waste to neutrons from the air. These neutrons are produced primarily by bombardment of oxygen and nitrogen nuclei in the atmosphere by cosmic rays and from spontaneous fission of natural isotopes in soil. Because Pu-239 is toxic in micro-quantities, this process, however slow, is potentially very dangerous. The radiotoxicity of Pu-239 is 3.4 x 10u times that of U-238. Consequently, for -4552 tons of uranium released in 1990, only 2.2 milligrams of Pu-239 bred by natural processes is necessary to double the radiotoxicity estimated to be released into the biosphere by year. Natural processes to produce both Pu-239 and Pu-240 appear to exist. For the 100 years following 1937, world use of coal as a heat source for electric power generation will result in the distribution of enormous radionuclides into the environment. This is very important problem and it is questionable about the risks and benefits of coal combustion, as source of electricity production. The potential health effects of released naturally occurring radionuclides are a long-term issue that has not been fully solved. Even with improved efficiency in retaining stack emissions, the removal of coal from its shielding overburden in the earth and subsequent combustion releases large quantities of radioactive materials to the surface of the earth. The emissions by CFPPs of greenhouse gases, a vast array of chemical by-products, and naturally occurring radionuclides make coal much less desirable as an energy source 21" RHD Jasná pod Chopkom 41 than is generally accepted. Large quantities of uranium and thorium and other radioelements in coal ash are not being treated as radioactive waste. These products emit low-level radiation, but because of regulatory differences, CFPPs are allowed to release quantities of radioactive material that would provoke enormous public outcry if such amounts were released from nuclear facilities. Nuclear waste products from coal combustion are allowed to be dispersed throughout the biosphere in an unregulated manner. Collected nuclear wastes that accumulate on electric utility sites are not protected from weathering, thus exposing people to increasing quantities of radioactive isotopes through air and water movement and the food chain [8]. The feet that large quantities of uranium and thorium are released from CFPPs without restriction increases a paradoxical situation. Considering that the nuclear power industry has been compelled to invest in expensive measures to greatly reduce releases of radionuclides from nuclear fuel and fission products to the environment, should coal-fired power plants be allowed to do so without constraints! Acknowledgement: Support from the Foundation Curie, Bratislava is gratefully acknowledged. References 1. World Resources Institute: World Resource 1996-1997. The Urban Environment. New York, 1996. 2. BP Statistical Review of World Energy 1998: Tables. 3. IAEA: Electricity, Nuclear Power and the Global Environment - Fact Sheets, 1998. 4. ABB Atom AB, Sweden, 1998. 5. Blix, H.: Fossil Fuels and the Environment. Joint IAEA/CNNC Seminar on 21st Century Nuclear Energy Development in China. 23 May 1997, Beijing, China 6. Ministry of Environment of the Slovak Republic: The first National communication on climate change. Bratislava, 1995. 7. Mr. Daniel, Uranpress Ltd., Spišská Nova Ves, private communication. 8. Gabbard, A.: Coal combustion: Nuclear resource or danger. ORNL, 1997. 9. McBride, J.P.; Moore, RE.; Witherspoon, J.P.; Blanco, R.E.: "Radiological impact of airborne effluents of coal and nuclear plants", Science. Dec. 8,1978. 10. Bertel: Nuclear energy and environmental debate, IAEA Bull., 37, 141 (1996) 11. U.S. Geological Survey: Radioactive elements in Coal and Fly Ash: Abundance and Environmental Significance. Fact Sheet FS-16397, October, 1997. 12. Taylor, M.: Greenhouse gases and nuclear fuel cycle: What emissions? IAEA Bull., 39, 111 (1997). 13. National Council on Radiation Protection: Public radiation exposure from nuclear power generation in the U.S., Report No. 92,72-112 (1987), 14. National Council on Radiation Protection: Radiation exposure of the U.S. population from consumer products and miscellaneous sources, Report No. 95, 32-36 and 62-64 (1987). 15. Blackburn, R.; Gueran, J.: Rad. Phys. Chem., 13, /3-4/145-147 (1979). 42 SK98K0337 21* RHD Jasná pod Chopkom THE NUCLEAR REGULATORY AUTHORITY OF THE SLOVAK REPUBLIC - INFORMATION TO THE PUBLIC Mojmír Šeliga Úrad jadrového dozoru SR, Bajkalská 27, 820 07 Bratislava A clear communications policy is the key to credibility and credibility is earned, not created. It is based on perceptions which give rise to varying levels of confidence. It has been consistently found in opinion research that credibility is the single most powerful persuasive force. Public communication programmes are the principal currency for the Regulatory Authority to inform the public on issues of cost, benefit, need and risk. For each issue the information needs differ and this must be reflected in the Regulatory's Authority communication programmes. The important aspect is testing if the nuclear energy in the Slovak Republic is due to obligatory rules acceptable and its operation is regulated by the state through the independent institution - The Nuclear Regulatory Authority of the Slovak Republic (UJD). UJD considers the whole area of public relations an essential component of its activity. UJD intends to serve the public true, systematic, qualified, understandable and independent information regarding nuclear safety of nuclear power plants, as well as regarding methods and results of UJD work. Generally, public information is considered as significant contribution to the creation of confidence into the regulatory work. The communication programme in the UJD The public relations are understood as attempts to establish, keep and improve UJD's good relations to its neighbours through purposeful informing. The UJD already in its origins laid the foundation of a policy of keeping the public broadly informed on the UJD activities and the safety of nuclear installations in the Slovak Republic by opening the UJD Information Centre. Catering to public & media relations, the Information Centre is instrumental in forming among the public a favourable picture of independent state supervision on nuclear safety. Professionally, the public relations at the UJD are under responsibility of the Public Information Manager, who is at the same time the press officer of the UJD . Of course, his close co-operation with all staff members is absolutely necessary. The Manager co- ordinates all public relations activities, but he also personally prepares press releases, writes articles, organises press conferences and communicates with TV, radio and journalists. He also monitors news in various media on subjects interesting for the UJD . An Information centre (IC) at the offices of UJD was built and opened in October (1995) with IAEA Director General Dr. Hans Blix as the first visitor. The entrance to the office building has been rebuilt and two rooms have been reserved for information purposes. The bigger room contains the Information centre, equipped with all electronic equipment. The room is big, enough to be used as a meeting room. The press conferences are held there too. I think that the Information centre could be a good tool to spread information to schools and interesting groups (members of Parliament, governmental and non-governmental groups and journalists). Approaching the UJD by telephone or through mail should be convenient, and information must be available without requiring a lot of effort from public. The basic rules for communication with media and with public are as follows: * To inform the public about the activities of the UJD, about its responsibilities, about The Status of nuclear installations safety in the Slovak Republic * To provide prompt, clear and consistent information to the public on a nuclear event 21" RHD Jasná pod Chopkom 43 whenever and wherever they may occur * To facilitate an independent communication between UJD, media and the public. Internal communication within UJD In the UJD, good internal communication within the authority forms the basis of good external communication. This is especially important, as the UJD is located both in Bratislava and in Trnava sites. Day-to-day information exchange between the two sites had been established. Daily faxes are sent from the Inspection branch at Trnava to the headquarters office every day. They deal with the status of the operations at the NPP sites. E-mail is also used for internal and external information by computer networks. As to other improvements of internal information it was noted that especially all staff meetings are held at regular intervals. This is an excellent form of internal information and for discussing external information matters. Handling the media UJD devotes considerable effort to be visible in the press and TV in radio programmes. Good relations has been established with a Slovakia's News Agency: the Agency asks UJD regularly for information and disseminates it to journalists. The Press Officer and other employees at the UJD have written articles published in various papers and journals, including specialised journals on nuclear safety. The advantage of articles written by staff members is that you can decide how your message is forwarded without a journalist D filter". The UJD is utilising all opportunities to contact the media. The Press Officer is aware of the importance of enhancing the UJD profile as an independent, non- promoting authority. The role of the UJD as an objective authority is always stressed in the articles. Last and this year were important for UJD in public information, because WS Atkins Science & Technology (GB) provided two courses on briefing the media in the event of nuclear accident. The courses were held in the headquarters of the Regulatory Authorities in Prague and in Bratislava. There were organised as series of mock interview interspersed with presentations on techniques and discussions, followed by a mini-exercise what required the participants to prepare press statements and hold mock press conference. In this year was important the visits of the Health Safety Executive's Office, the British Nuclear Industry Forum, UK NEA, AEA Technology and other departments to provide an opportunity to be aware of the public information matters and the co-operation with the media in the United Kingdom and have the opportunity to speak with the people with the same interest about their way of dealing with the subject. I have gain a lot of information about the public relations and about the co-operation with the media in UK, which can be compared with ours as well as with those from our countries I have visited before and used for improving of my work in the UJD. It is important to confer not only information but also message. The structure of a release has to comply with journalistic work: "the most important information (message) comes first, the least important last". Sometimes, it is difficult to make a release, especially if the topic is very complex. In such a case, background material should be enclosed to the release. The basic rule for deciding who deals with the media on a particular issue is "who is the most appropriate person?". It should be someone who knows about the issue; who can speak with authority; and, who is used to dealing with the media. This will not necessarily be the most senior person available. Nevertheless, when dealing with important or high profile cases a senior officer will be needed because media representatives will expect to speak to "someone in authority". 44 2 ľ RHD Jasná pod Chopkom It is important that all UJD VIP staff coming into contact with the media: project a positive and professional image of UJD and its work deliver a consistent message - are familiar with all the facts of the situation which has aroused interest are sensitive to potential areas of misunderstanding are prepared - avoid speculation inform UJD's Press Office The media documents The media publication of UJD in the detailed annual report, written in Slovak, who has been sent to involved ministries, enterprises, nuclear installations and related industries and research institutes and related University. For public information purposes, nicely illustrated, in Slovak and English versions and has been prepared too. This version is sent to the Government, Foreign embassies in Bratislava, Embassies of Slovak Republic abroad and to other International organisations and the media. UJD has published a bulletin, which appears three or four times annually. It cover consists of current issues regarding nuclear safety of NPPs and local and international activities of the UJD. The UJD intends to send it to the media, politician's etc. The UJD has one issue translated into English. The relations with media have been prioritised in the UJD since beginning of activities in the Information Centre. Media release are often the only way to reach a broad public with limited resources. The disadvantage is that one cannot manage the message interpretation. It is therefore very important that the way the authority expresses its message is easy and understandable for the journalists. The UJD has direct access to the relevant information including reporting of events through its resident inspectors. Nevertheless, the main information channel to the UJD is provided directly by the NPP staff. Daily reports on the status of individual reactor units are sent to the UJD regularly every morning. The NPP is also primarily responsible for dissemination of first appropriate information about operational events to the public. The emergency preparedness and media The UJD has Emergency Task Group, which is divided into Reactor Safety Group, Radiation Protection Group, Logistic Group and Information Group. Information Group consists of the head of Group, the report analysis expert and news service expert. The other members are from the Emergency Preparedness Department and other staff members from UJD. The information group is responsible for public information as well as for informing the authorities. The Information Centre is situated close to the Emergency Response Centre; it intends to be used both during emergency drills as well as during real emergencies. Over the past year UJD has been working toward demonstrating its effective response to an emergency. Information group was in all exercises responsible for co-ordinating the technical briefing material prepared by the Reactor Safety Group and Radiological Assessment Group and was produced in every exercises briefing material for National Emergency Commission for Radiation Accidents, for domestic and international organisations. Also the dissemination to the public and media was tested during these exercises. The Information Group was also preparing messages which were sent to international bodies such as IAEA and, to neighbouring countries as a part of Slovak Republic bilateral arrangements. More detailed messages were sent to Parliament and selected Ministries of the Slovak Government. Furthermore, the Information group monitored reports issued by the media in TV and radio. 2 ŕ RHD Jasná pod Chopkom 45 The Activities of the UJD in last period Numerous missions of public relations experts from Sweden, Finland, UK., Germany and the USA helped to prepare and train UJD specialists, who manifested itself in a capability to arrange a number of seminars and meetings with schools and journalists and the periodic press conferences on major activities of the UJD. As co-ordinator of special projects of the public information division of the IAEA prepared UJD in March 1998 the seminar "Nuclear Safety and the Media". The UJD issues the Bulletin on UJD activities for domestic and foreign public. On the occasion of the 40th anniversary of the IAEA, we prepared a special publication on the IAEA, organized a visit to the IAEA encompassing by reception of its director general, H.Blix, of Slovak Member of the Parliament. At the end of the 1997 year an international meeting for foreign diplomatic corps in the Slovak Republic, senior officials of the Slovak state Administration, universities, research institutes and other state authorities and organizations along with Slovak nestors who were involved in founding the IAEA and elaborating fundamental agreements on peaceful uses of atomic energy the occasion of the 40th anniversary of the IAEA was organized. Seventy-five contributions (press releases) on UJD national and foreign activities were transmitted to the Press Agency of the Slovak Republic (TASR) over the course of 1998 (additionally, information was communicated to the reading and electronic media.) In the journal "Bezpečnosť JE (Safety of Nuclear Energy)" (bimonthly), the column "Information" periodically carries briefs on UJD activities (inspection activity, organizing seminars and expert meetings, important home and foreign visits.). Very important for the UJD activities was in the year 1998 the commissioning of the NPP Mochovce. Contributions on UJD regulatory activities and international co-operation are periodically put out in the "Slovenské elektrárne" Newsletter, the "Mochovce" and "Bohunice" in-house journal, the Slovak Nuclear Society Bulletin. Report on the Safety of NPPs in the Slovak Republic is annually published in the Journal European Nuclear Society - Nuclear Energy Worldscan. Some articles on UJD activities were published in the world information agency NucNet. The UJD issued in 1998 three editions of the internal Bulletin on the UJD's national and foreign activities and personnel. In this year were prepared 2 articles with photos documentation about the main functions, Status and activities of the UJD in connection with The World Energetic Conference in American Huston (September 1998) and special edition about the Slovakia, printed by the Spain Embassy (May 1998). The four press conferences were held in part of year 1998. Radio and TV was cover a press conference, but interviews with broadcast are usually recorded immediately after a press conference. Conclusions UJD as the state authority provides information related to its competence, namely information on safety of operation of nuclear installations, independently from nuclear operation and it enables the public and media to examine information on nuclear installations. More active public information activities of the UJD will significantly contribute to the public understanding on different aspects of the uses of nuclear energy and will increase the public treats in this area. There is still a worldwide opposition against nuclear energy but not in an amount as before 5-10 years. Using most modern and effective tools like the internet but also by presenting different and high quality materials and publications the IAEA and other international agencies could accelerate the process of public's positive attitude with respect to nuclear energy. During this year Slovakia hosted a seminar of this kind which was a great success. 46 21"RHD Jasná pod Chopkom A further part of the reform process is the ongoing review of the role and management of public information. The importance of public understanding of the role of nuclear energy and of the UJD demands that effective public information be an integral part of the UJD's activities. Very important is the opinion of the IAEA Director General, that the size of the Agency's public information programme is modest (The Board of Guvernors, Vienna September 14- 16, 1998). It is therefore all the more essential that it be effectively targeted, particularly towards the nuclear, arms control and development communities, using the most modern and effective tools. Over the coming months I will be exploring Member State interest in an expended programme of public information seminars obvious, however, that a lot of work is still to be done, especially with the aim to assure better direct presentation of UJD to the public as a competent and independent regulatory organisation. All UJD communication and information activities in 1998 presented here in aim to Creation of public confidence, favourable UJD image at home as well as abroad. 2 ŕ RHD Jasná pod Chopkom SK98K0338 47 PREČO A AKO KOMUNIKOVAŤ S VEREJNOSŤOU O RADIAČNOM RIZIKU RNDr. Soňa Ftáčniková, CSc, Ústav preventívnej a klinickej medicíny, Bratislava ÚVOD Verejnosť je značne znepokojená keď sa hovorí o ionizujúcom žiarení a o radiačnom riziku s ním spojeným. Existuje množstvo štúdií z rozličných častí sveta, z ktorých sú evidentné obavy laickej verejnosti a nedôvera k oficiálnym vyhláseniam a opatreniam hlavne v súvislosti s výrobou elektrickej energie v jadrových elektrárnách, s uskladňovaním jadrového odpadu, špeciálne po nehode v Černobyle [1,2]. Táto nedôvera má svoje korene v zlyhaní vzájomnej komunikácii medzi odborníknii a verejnosťou. Tým, ktorí majú vedomosti (odborníkom) sa neverí, že hovoria úplnú pravdu alebo, že im záleží na bežnej populácii a na druhej strane odborníci vidiac, že vo verejnosti sa objavujú ničím logicky nepodložené emócie, neveria, že laická verejnosť je schopná pochopiť odbornú informáciu o potenciálnom riziku. Preto je nevyhnutná efektívna, plodná a otvorená komunikácia, schopná prelomiť túto vzájomnú nedôveru. DEFINÍCIA RIZIKA Jednou z najdôležitejšou zásadou je jasne, zrozumiteľne a najväčšou možnou presnosťou zadefinovať a vysvetliť pojmy o ktorých sa vedie diskusia. Sadman (5) zadefinoval risk ako: RISK = HAZARD + SOCIÁLNY PARAMETER (OUTRAGE) HAZARD = technické riziko = [pravdepodobnosť výskytu udalosti počas určitého časového obdobia] SOCIÁLNY PARAMETER (OUTRAGE-násilie) - odráža kvalitatívnu stránku rizika, súvisí s „nevedeckými faktormi", ktoré môžu zvýšiť alebo znížiť vnímanie rizika a tiež súvisí s hodnotovým rebríčkom spoločnosti teda s etickými normami danej komunity (aj keď technické riziko je nulové, stále tu môže existovať risk spôsobený druhým členom vzťahu = sociálnym parametrom). RISK PERCEPTION (ako je pociťované, vnímané riziko a jeho relatívna dôležitosť na život členov spoločnosti). Podľa Sandmana (5), odborníci sa často koncentrujú výlučne na určovanie hazardu, na rozdiel od komunít, ktoré sa v extrémnych prípadoch sústreďujú výlučne na „outrage". Toto potom vyúsťuje do častých nedorozumení, k nezhodám vo vzájomnej komunikácii, následným konfliktom a nedôvere. Kľúč k úspešnej komunikácii je v nezanedbaní a podcenení ani jedného z komponentov v definícii risku. Niektoré v literatúre uvádzané „outrage,, faktory v priamom premietnutí na riziko spojené s využívaním ionizujúceho žiarenia - katastrofické následky (atómová bomba, nehody v elektrárnách) - smrteľné následky (choroby z ožiarenia, rakovina) - neviditeľné riziká (žiarenie) - nedobrovoľné riziká (iné ako medicínske využitie žiarenia) - nedôveryhodné zdroje (Černobyl) - nekontrolovateľné riziko (odpady v životnom prostredí) 48 2 ľ RHD Jasná pod Chopkom - neetické riziko (odpady v život, prostredí = environmental unjustice) - neznáme riziko (stochastický, „non threshold effect") - umelo vytvorené riziká (atómové elektrárne, bomby) Tabuľka 1 obsahuje informácie o tom ako vnímajú rozličné riziká v našom životnom prostredí experti a ako verejnosť. "Outrage" Hazard Vysoký Nízky Vysoký Otrava olovom v detsve Skládky jadrového odpadu Pasívne fajčenie Ožarovanie potravín Nízky Expozícia radónom Aflatoxíny Patogény v potravinách Chlórovanie vody ACCEPTABILITY OF RISK (ako verejnosť akceptuje, prijíma riziko) Obavy o zdravie a kvalitu života sú úzko spojené s pocitom možnosti kontroly nad svojim životom. Ľudia oveľa ľahšie akceptujú riziko, ktoré podstupujú dobrovoľne a majú nad ním kontrolu (fajčenie, radón v domoch, doprava, nebezpečné športy) aj keď je niekedy oveľa vyššie ako riziko nad ktorým nemajú kontrolu a podstupujú ho nedrobrovoľne. Tabuľka 2: dobre akceptovateľné ťažko akceptovateľné riziko riziko dobrovoľné nedobrovoľné chronické náhle, katastrofické bežné riziko smrteľné následky okamžitý efekt neskorší efekt známy účinok neznámy efekt expozícia riziku je nutná expozícia je luxus riziko zasahuje informovaných citlivú skupinu (deti, dospelých handikepovaných) následky sú reverzibilné ireversibilné Starr's Laws (1969) o akceptovateľnosti rizika - verejnosť je ochotná akceptovať riziko z dobrovoľných aktivít asi 1000 krát vyššie ako z nedobrovoľných aktivít - akceptovateľnosť výšky rizika je nepriamo úmerné počtu exponovaných jedincov - výška rizika dobrovolné tolerovaného je na úrovni rizika choroby acceptibility of risk O [benefiet]3 Tabuľka 3: Dobrovoľne podstúpené riziko, ktoré zvýši pravdepodobnosť smrti o faktor Iff6 ( 1 z milióna) (4) Aktivta Príčina smrti fajčenie 1,4 cigarety denne rakovina, kardiovasculárne ochorenia vypitie" 1 vína cirhóza pečene 21st RHD Jasná pod Chopkom 49 prežitie 2 dňov v New Yorku znečistenie vzduchu prežitie 2 mesiacov v Denveri rakovina z kozmického žiarenia bicyklovanie, asi 15 km nehoda lietanie lietadlom, asi 1500 km nehoda jeden RTG snímok hrudníka rakovina zo žiarenia zjedenie 1 pohára arašídového rakovina pečene spôsobená masla aflatoxínomB 1 ročné pitie vody z vodovodu rakovina z chloroformu KOMUNIKÁCIA O RIZIKU Nikto z nás nepochybuje, že je nutná a užitočná komunikácia medzi odborníkmi a verejnosťou ( ľuďmi, ktorí sú jednak konzumentami produktov pri ktorých sa vytvára radiačné riziko a tiež sú „obeťami" žijúcimi v prostredí poznačenom touto produkciou). Boli vypracované rôzne scenáre a návrhy založené na konkrétnych skúsenostiach o tom ako úspešne a efektívne komunikovať. Je nevyhnutné dodržiavať tri základné princípy a síce: byť proaktfvny ( ako opak k reaktívnemu), komunikácia musí byť proces nepretržitý a kooperatívny. Najdôližitejšie pri tvorbe stratégie úspešnej komunikácii je stanoviť si realistický cieľ = podpora zvyšovania úrovne vedomostí a chápania konkrétneho rizika, jeho možných riešení všetkými zainteresovanými. Russell vo svojej práci (6) píše, že o úspechu v komunikácii o riziku môžeme hovoriť vtedy, keď si dobre informovaná verejnosť sama vyberie opatrenia znižujúce riziko, tak ako ich navrhnú odborníci a nie keď len prijme tieto riziko minimalizujúce návrhy. Ako veľmi účinné a prospešné v zlepšení vzájomnej komunikácie sa ukázali niektoré dôležité zásady pre odborníkov komunikujúcich o radiačnom (ale nie len) riziku: 1/ buď čestný, úprimný a otvorený v komunikácii o riziku a o alternatívnych možnostiach riešenia 2/ akceptuj a zainteresuj laickú verejnosť ako legitímneho partnera 3/ počúvaj pozorne druhú stranu a skús sa vžiť do jej situácie a pocitov 4/ nevyhýbaj sa médiám 5/ hovor jasne, nepoužívaj príliš odbornú terminológiu 6/ používaj vhodné ( objektívne ako je to len možné) prirovnania jednotlivých rizík a udávaj aj veľkosť neurčitosti pri určovaní pravdepodobnosti uskutočnenia sa danej udalosti 7/ jasne vysvetli rizikovú situáciu so všetkými možnými následkami s príslušnými pravdepodobnosťami zrealizovania danej udalosti a uveď či prezentované čísla sú objektívne zistené pravdepodobnosti výskytu udalosti, alebo sú tvoj osobný názor 8/jasne a otvorene vysvetli dôvody, ktoré viedli ku konečnému riešeniu na verejnom stretnutí s druhou stranou 9/ vysvetli ako "manažovať" (riadiť) riziko. Napriek tomu je treba mať neustále na pamäti, že každá situácie je svojim spôsobom jedinečná a neexistuje univerzálny návod, ako komunikovať o vzniknutom radiačnom riziku. Dôvera k autoritám, odborníkom a informáciám nimi poskytnutými informáciám ( bod 1) je kľúčová pri vnímaní rizika. Okamžité celosvetové negatívne reakcie po Černobyle, vo vzťahu k nukleárnej energii, postupne zmizli a situácia sa ustálila na úrovni pred nehodou. Dnes dôvera k informáciám o úrovni rárioaktivity v životnom prostredí je značne vysoká. Obr. 1 ilustruje, že 75 % populácie žijúcej v Európskej spoločnosti má „úplnú" alebo „vysoký stupeň dôvery" k nezávislým vedcom ( podľa 50 2ľ'RHD Jasná pod Chopkom údajov Commission of European Communities 1991 (3)). Iná situácia je v prípade dôvery k verejnej správe ako je to zrejmé z grafu na obr. 1. TAK TEDA AKO S IMPLEMENTÁCIOU ÚSPEŠNEJ KOMUNIKÁCIE O RIZIKU ? Ako sme už poznamenali v úvode úspešná a efektívna komunikácia sa musí opierať o tri základné princípy a síce: musí byť proaktívna, nepretržitá a kooperatívna. Pod pojmom proaktívna myslíme, že pri komunikácii o riziku sa nesmie čakať, keď sa už nejaká udalosť stane a potom reagovať vysvětlováním a opatreniami. Je nutné dopredu "proaktívne" oslovovať rozličné skupiny a nepretržite vysvetľovať a vzdelávať a to v úzkej spolupráci a priamom zainteresovaní oslovených komunít. Čo sa týka komunikácii konkrétne o radiačnom riziku, úlohou číslo 1 je zaviesť do škôl prvého aj druhého stupňa vhodný rádioekologický vzdelávací program, s cieľom osloviť mladú generáciu a už od mala podávať objektívne pravdivé, emóciami neskreslené informácie = proaktívny prístup a nepretžitý proces. Ďaľší veľmi účinný spôsob komunikácie je vzdelávanie komunity nezainteresovanými vysokokvalifikovanými odborníkmi. Ide predovšetkým o kurzy venované problému „risk assesment" pravda pri dodržaní zásad vhodnej komunikácie ako boli uvedené vyššie. Tento vzdelávací proces musí byť správne cielený, nepretržitý, intenzívny a trvalý, aby bol účinný = proaktívny prístup a nepretžitý proces. Kooperatívnosť = vytváranie programov na interkomunikáciu medzi odborníkmi a komunitou. Jedným z takýchto príkladov je úsilie amerického DOC ( Department of Energy) o efektívnu a plodnú komunikáciu v súvislosti s hnutím ochranárov životného prostredia. Ochranárom boli fiktívne poskytnuté peniaze, ktoré by si vyžadovalo vyčistenie jednej oblasti od zbytkového rádioaktívneho odpadu s aktivitou pod prípustným limitom pre životné prostredie. Potom boli aktivisti vyzvaní vytvoriť priority v použití poskytnutých peňazí. Mnohí aktivisti kompletne zmenili priority, keď boli postavení pred rozhodnutie použiť peniaze na komplexné riešenie existujúcich problémov v danej oblasti. Teda je nevyhnutné zahrnúť lokálne komunity do rozhodovacieho procesu. V danom projekte sa ukázalo, že tieto komunity po vzájomnej komunikácii s odborníkmi sa rozhodli vo väčšine prípadov využiť tieto slabo kontaminované oblasti na rekreačné a priemyselné účely a ušetrené peniaze použiť na kompletizáciu kanalizácie, ktorá sa v danej dobe a oblasti bola oveľa pálčivejší problém pre komunitu. LITERATÚRA 1/ Brenot, J., Hessler, A., Joussen,W., Sj5rberg,L: Perception of radiation risk from a cross cultural perspective. Book of Proceedings from IRPA 9, 1996, vol.4., pp. 4-790 - 4-792. 2/ Drottz-Sjorberg, B.M., Persson, L: Public reaction to radiation: Fear, anxiety, or phobia ? Health. Phys., 64(3): 223-231, 1993 3/ Commission of the European Communities. Eurobarometer: Public opinion in the European Community. Brussels, Directorate- General Information, Communication, Culture, no.35, June 1991 4/ Wilson,R.: Technology Review, Feb. 1979, p.45 5/ Sandman, P.M.,: Risk Communication: Facing Public Outrage, EPA Journal, 13, 1987, pp.21-22. 2 ľ RHD Jasná pod Chopkom 51 6/ Russell, M.: Risk Communication: Informing Public Opinion, EPA Journal, 13, 1987, pp.20-21. Obr.l: Dôvera verejnosti v EC v rôznymi skupinami poskytovanými informiciami o úrovni rádioaktivity v iivotnom prostredi percento dôverujúcich O percento nedôverujúcich Nezávislí Universitní vedd profesori a učitelia 52 SK98K0339 21st RHD Jasná pod Chopkom IMPROVEMENT OF INFORMATION ON THE NUCLEAR ENERGY HEALTH EFFECTS, THE AIM OF WIN SLOVAKIA Petrášová, M.1}, Nikodémova D.2), ^Nuclear Power Plants Research Institute Trnava Inc., 'Institute of Preventive and Clinical Medicine Bratislava Introduction The methods of the radiation risk communication to the general population is not always on the sufficiently qualified level in our country. This can be the cause of the lack of interest, even opposition and fear among the population, including the young generation as far as the nuclear energy and the application of the radiation sources is concerned. After all, this is not a specific problem of Slovakia. It concerns other countries too, as it is demonstrated by a great number of articles and studies devoted to the perception of the radiation, to the assessment of the communication methods and to the methods of comparison with the potenciál risks from other sources.(air-craft accidents, AIDS, chemical pollution etc). In our contribution we would like to summarize some known results cocerning the communication of the risk probability and the possible health effects of radiation to laymen based on the serious conclusions of the epidemiological studies and on the modeling of the various situations. Activities WIN Global and WIN Slovakia There are very few institutions which profesionally deal with the problem of the stochastic risk comparison as a basis for more effective communication in the nuclear energy. From this point of view we consider the role of women and young generation in the radiation risk communication as very essential and important. In this respect international organisation WIN Global and our national organisation WIN Slovakia (women in nuclear) which as a section of Slovak Nuclear Society, offer unique oportunities for the improvement of radiation risk communication. WIN Global was established in 1993 and currently has about 600 members in 39 countries. Our national WIN Slovakia was established in the end of 1997 and has 20 members. WIN Slovakia is the association of women working professionally in the fields of nuclear energy and application of radiation and willing to devote time to public information. While most of the members of WIN are employed in nuclear energy sector, there are some members working in the other areas, where nuclear technologies are utilised. The membership includes for example women working in medicine and health care, in regulatory authorities, industry and the independent researchers. We are sorry there are no members in WIN Slovakia from the university intitutions so far. WIN is open to men supporting the goals of WIN. WIN's principal objective is to emphasise and support the role that women can and do have in addressing the general public's concerns about nuclear energy and the application of radiation and nuclear technology. WIN can do this through educational 2ľ RHD Jasná pod Chopkom 53 programmes, information exhange and arranging study visits. There is no difference among the goals of our national organisation and WIN Global. Members of WIN Slovakia all have one thing in common: They want the general public to have a better understanding of nuclear and radiation matter. The members of our organisation would like and plane to make presentations, discuss and give information material on subjects as: energy and sustainable development radiation, radioactivity, and health effects medical applications, radiation protection nuclear energy, uranium mining nuclear power plants and their safety radiaoctive waste nuclear and environment natural radiation, radon Risk perception and some results of research in Slovakia Nuclear power is a questionable source of energy, because of the perceived risks connected with it. Acceptance of nuclear technology depends on the assessment of both advantages and disadvantages. Underlying only the benefits is not enough. It is very important to explain the risk too. One of the most important methods of communication for improving the risk information among the population is themethod of mutual comparison of the various types of risks. People perceive usually risks and behave accordingly without knowing the real risk. An understanding of people perception of risk and factors that influence the risk perception is the essential basis for risk communication. Radiation is an emotially involving issue, which avokes a wide range of reactions and arguments. To be accepted a risk management policy must be transparent and trustworthy. Much have been written about social and cultural influences on construction of radiological risk and the need of information and education of the people. The best way to understand the public's view of various applications of ionizing radiation is to ask people directly by means of interview and surveys In 1996-1997 a comparative risk perception study was carried out in Slovakia[l]. Real data were collected through the administration of a questionaires distributed among a group of 14-17 years old children (Ni = 308) and teenagers (N2=150). Respondents were asked to indicate the degree of health risk associated with various types of risks (tabl) [2]. This list of 44 items covered a wide range of risks and hazards, including risks from technology ( nuclear power plants, waterdams etc.) polution (air-, water-, soil, waste management) nature (floods, fire, etc.), life style (smoking, drugs, alcohol abuse) and society (crime, conflicts, war, terror etc.) The questionaire contains the questions about the sources of risks information. The topic of the study was the self assesment of the knoledge on particular risks too (four point scale - very good, good, weak, no knowledge) The results can be summarized in the following conclusions: 1) Hierarchy of the perceived risks has shown, that as the most threatening ones at present are the risks associated with nuclear power ( radioactive waste, radioactivity, nuclear power plant) and the degradation of environment (air-, water-, soil- pollution, chemical in food). Smoking, alcohol and medicine abuse, noise etc. received less attentioa 54 21st RHD Jasná pod Chopkom 2) Media seem to be the most frequent information resource on most of the risks in the group of 14 -17 years old children as well as in the group of teenagers. The weekest risks information resource seems to be the school. 3) Both children and teenagers evaluated their own knowledge on particular risks as good, or weak. But the most frequent answers indicating the level „ no knowledge" were conected with the risks in the nuclear power, social and sports related risks. Tab. 1: List of various risk and threats 1. Criminality (violence, robbery, mafia) 23. Alcohol 2. Mountaineering 24. Conflicts among people 3. Traffic accidents 25. Medicine abuse 4. Smoking 26. Overcrowding 5. Unemployment 27. Radioactivity 6. Radioactive waste from nuclear 28. Swimming power station 29. Soil pollution 7. Skiing 30. Cancer 8. Drug abuse 31. Gambling 9. Computer games 32. Diseases due to the effect of the environm. 10. Divorce 33. Extreme groups (skinheads...) 11. Water pollution 34. Woterworks Gabčíkovo 12. Cycling 35. Chemical substances in food 13. Terrorism 36. Floods 14. War conflict 37. Lack of money 15. Football 38. Dams 16. Nuclear power station 39. Epidemics (typhus, plague) 17. Political instability in the country 40. Loss of relatives and friends 18. Dams 41. Lack of food 19. Noise 42. Fires 20. AIDS 43. Animal-transmitted diseases? 21. Unwanted parenthood 44. Loneliness 22. Air pollution Conclusion On the basis of this research and similar studies there exists evidence that if we want to put the radiation risk into perspective we have to improve the risk communication among the specialists and laymen. The campaign for the improvement of the risk communication must be started in schools. After the Chernobyl accident Ministry of health elaborated the special programme of education concerning the nuclear energy and radiation risk in regular education process for the primary and secondary schools, but this programme has not been realized so far. It is urgently recomended to return to the educational programme and put it in force in an innovated form. We have to adress the people to create the open discussion of risks, perception and valuation which will lead to understanding and mutual confidence. From this basis it is possible to add the information about benefits, needs and other possibilities as environmental advantages, sustainability and responsibility 21" RHD Jasná pod Chopkom 55 By the communication with female groups having concerns about nuclear, the risks has to be explained and put in perspective. It is even possible to reach the point, where nuclear will be judged as one of the green alternatives. Literature 1 Rosová,V., Rošková,E., Bianchi,G.: Vnímanie rizík v detskej populácii na Slovensku, AEUC 1996, 8, s.76-86 2 NikodemovájD., Rosová, V.: Some remarks on public perception of the risk from Radiation in Slovakia. Proc. of International Symposium on Environmental Epidemiology in Central Europe, Smolenice, Sept.24.-27., 1997 56 21st RHD Jasná pod Chopkom SK98K0340 RADIONUCLIDE COMPOSITION OF NOBLE GASES IN EFFLUENTS OF SOME NUCLEAR ESTABLISHMENTS Rulík P., Tomášek M., Malátovál. National Radiation Protection Institute, Prague, Šrobárova 48, Czech Republic Introduction Radioactive noble gases are important part of effluents from ventilation stacks of nuclear power plants and other nuclear reactors. Composition of the mixture varies in dependence on the technology of the ventilation, operation mode of the reactor (normal operation, phase of the shut-down of the reactor) and on the quality of fuel elements cladding. In the frame of the long-time co-operation there are regularly monitored by NRPI the gas effluents fromNPP Dukovany and from ventilation stack of the reactor in Nuclear Research Institute Řež (ÚJV Řež). The method of measurement and sample analysis The method is based on the collection of compressed air taken from the ventilation stack in 6 pressure flasks, each of which has a volume 0,35 L and the immediate gamaspectrometry measurement using semiconductor HPGe detector. The sampled air is filtered to remove aerosols and then compressed to pressures from 19 to 20 MPa. The pressure flasks are connected in parallel and form a ring (Fig.l) into which the detector is inserted [1]. In such a geometry the best efficiency and reproducibility of measurement is achieved. The sampling requires about 8 minutes. The sample measurement is usually performed in the measuring room near the sampling place without shielding. The time of measurement is from 10 to 60 minutes. The last samples are taken to NRPI Prague for second longer measurement in the shielding. Calibration of the HPGe detector for the described geometry of measurement was carried out using one flask filled with homogenous mixture of standard solution of radionuclides, small spheres of foamy polystyrene and epoxy resin. The density of this mixture was same as the density of pressured air with noble gases. Gamma spectra measured by a portable HPGe detector with relative efficiency 15 or 35 % are accumulated in a multichannel analyser Canberra Series 10+ (Fig. 2) and stored on a harddisk in a notebook for later evaluation in the laboratory. In the last sample from ventilation stacks were also determined 85Kr in laboratories at Prague. The method of 85Kr determination is based on the separation of krypton from sample by cryogenic adsorption on beds of active charcoal arid on the radiometry of 8 Kr by CaF2(Eu) scintilation detector. Afterwards, analysis of the separated sample for krypton element on a gas chromatograph aimed at the separation efficiency evaluation is performed. Results In the paper there are shown the results of measurement from the ventilation stacks of NPP Dukovany and from the ventilation stack of reactor ÚJV Řež in the time period 1996 - 1998. These results are compared with the results of similar former measurement from NPP Jaslovské Bohunice and from NPP Dukovany [1]. 2 ľ' RHD Jasná pod Chopkom 57 The minimum detectable activities (MDA) for evaluated radionuclides at 95% confidence level are in table 1. It is necessary to have in mind that the MDA for the most of noble gases will be in real higher due to higher Compton background from 41Ar and other presented radionuclides in a real sample. Only 4IAr, 85Kr, 133Xe and I35Xe are measureable at NPP Dukovany in these conditions by this method at present time. Their average activities are shown in table 2. Table 1: Minimum detectable activities of evaluated radionuclides at confidence level 95% for measuring time 20 minutes without shielding in the room near ventilation stack and iron shielding with 200 mm wall thickness in the laboratory for HPGe detector with relative efficiency of 35%. Nuclide Half life MDA [Bq.m'^l Without shielding with shielding Ar 41 llOmin 20 2 Kr85 10,7 y - 0,4 * Kr85m 4,5 h 30 2 Kr87 76 min 40 3 Kr88 2,9 h 50 4 Xe 131m ll,8d 1100 60 Xel33 5,2 d 200 14 Xe 133m 2,2 d 200 10 Xel35 9,1 h 20 2 Xe 135m 15,7 min 50 6 Xel38 14,2 min 80 6 by the method for evaluation of 'Kr Table 2: Average activity concentration (in Bq/m3) of 41Ar, 85Kr, I33Xe and 135Xe in effluents from ventilation stacks (VS) of NPP Dukovany during normal operation in time period 1996- 1998. Nuclide VS- 1 VS-2 Bq/m3 Bq/mJ Ar41 800 660 Kr85 34 17 Xel33 <60,170 > <20,210> Xel35 <40,60> < 0,20 > < > .. .range of values (many of the individual values are below MDA) Conclusions This method can be recommended as a very useful independent method of monitoring of noble gases effluents from ventilation stacks of NPPs. When used at routine NPP operation for intermittent monitoring with regular intervals - e.g. as a complementary to a continuous gamma spectrometry of noble gases - the pressure vessel is supposed to be taken to a good laboratory shielding rather than to place a portable spectrometer near to the ventilation stack. 58 2 ľ RHD Jasná pod Chopkom References [1] Malátová L, Rulík P., Bučina L, Kuča P.: Method of analysis of a mixture of radioactive noble gases in airborne effluents from nuclear power plant, Bezpečnosť jadrovej energie, 1(39), 1993,4 Fig. 1: Six pressure flasks connected in parallel and form a ring VS-1 VS-2 unsssjs 88&SS888 8 S & S S S S 5 •«;•:•;« tó^OOWOÔÔi w » Ô ó d d ä ä M o o d Date of sampling Date of sampling Fig.3 Average activity concentration of Ar Fig.4 Average activity concentration of Kr 41 in the individual sampling days from 85 in the individual sampling days from ventilation stacks (VS) of NPP Dukovany ventilation stacks of NPP Dukovany O - normal operation S - shut down of one reactor (each ventilation stack is common for two reactors ) 21"RHD Jasná pod Ckopkom SK98K0341 RADIOTRACER STUDY OF THE ADSORPTION OF Fe(m), Cr(III) AND Cd(II) ON NATURAL AND CHEMICALY MODIFIED SLOVAK ZEOLITE Mária Fôldesová, Pavel Dillinger, Peter Lukáč Faculty of Chemical Technology, Slovak University of Technology, Radlinského 9, 812 37 Bratislava, Slovak Republic In order to minimize the contamination of environment with metals in ionic form the more types of natural and chemicaly modified zeolites were examinated to their uptake of Fe(III), Cr(III) and Cd(II) from water solutions by batch radioexchange equilibration method. Excellent and relatively cheap inorganic ion exchanger - zeolite - offer the advantages of high capacity, good selectivity, reproducible stoichiometry, radiation resistance and compatibility with glas and cementation waste forms. The fundamental building block of the zeolites is a tetrahedron of four oxygen atoms surrounding a relatively small silicon or aluminium atom. The structure consists of SÍO4 and AIO4 tetrahedra arranged so that each oxygen atom is shared between two tetrahedra. Because aluminium has one less positive charge than silicon, the framework has a net negative charge of one at the site of each aluminium atom and is balanced by the exchangeable cation. The cavities and channels in this structure can hold metal cations (mostly alkali metals and alkaline earth metals) or molecules (e.g. H2O). The structure of a zeolite determines its specific physical and chemical properties, such as ion-exchange ability and adsorption, reversible dehydration and hydration. The chemical treatment of zeolites with excess NaOH is excepted to reduce the effect of naturally occuring cations on their occlusion. The changes in the structure of zeolite in general increase its sorption properties mainly for multivalent cations [1,2]. In this study was used zeolitic tuff from deposit NiDný Hrabovec (content of clinoptilolite 50 -70 %) with the grain size from 1,2 to 2,2 mm. The priority of Slovak zeolite is the absence of mordenite, which has a fibrous structure and can be regarded as carcinogen material. The granulas of zeolite were modified wih the folloving NaOH solutions: 0,5, 1, 2 and 4 mol.ľ1 at 80 °C for 4 hours [3]. For the determination of equilibrium time period zeolite was mixed with 15 ml 1 59 51 0,05 mol.1" FeCl3.6H2O or Cr(NO3)3.9H2O or CdCfc solutions labeled with Fe or Cr or 115mCd in glass bottle via different ratios of mass zeolite (50, 100, 150, 200, 250, 300 mg). The mixture was gently shaken during 7 days. The 5 ml of solution was decounted for gama-ray counting by Nal/Tl/ detector every 24 hours. The sorption of Fe, Cr and Cd on all types of zeolites was studied by radioexchange method and the sorption of Fe and Cr also by flame atomic absorption method. From sorption curves were the sorption coefficients calculated. The adsorption of Fe(HI), Cr(III) and Cd(II) on the zeolites modified with NaOH increases with the concentration of NaOH solution. For the zeolites modified with 4 mol.1"1 NaOH solution the adsorption value for Fe and Cr was about 30 % that is about six time increase, and for Cd about 50 % that is about two time increase as compared with the natural zeolite. 60 2 ľ RHD Jasná pod Chopkom The correlation between the radioexchange method and flame AAS method was good, in the range from -4,2 to +3,2 %. The results obtained in this work show that zeolites modified with NaOH solution are suitable for adsorption of Fe(IH), Cr(HI) and Cd(II) from underwater, waste water, feed water and coolant water from nuclear power plants. The adsorbed zeolites can be solidified by convention way. Nontheless, each specific problems should be examined individually, for in solution as heterogenous as natural waters and radioactive waste waters, unpredictable specific interactions of the numerous factors involved in this process frequently lead to unexcepted results. REFERENCES 1. Lukáč P., Fôldesová M.: Sorption properties of chemicaly treated clinoptilolites with respect to Cs and Co XRadioanal.Nucl.Chem., Letters 188/6/,427-437(1994) 2. Godelitsas A., Misaelides P., Charistos D., Filippidis A., Anousis I.: Uranium sorption from aqueous solutions on sodium form of HEU-type zeolite crystals J.Radioanal.Nucl.Chem., Letters 208/2/,393-402(1996) 3. Mozek P., Fôldesová M., Lukáč P.: Study of NaOH treated clinoptilolites and their physical and ion-exchange characteristics J.Radioanal.Nucl.Chem., Letters 165/3/,175-183(1992) 2ľ' RHD Jasná pod Chopkom I llllllll lllll III lllll lllllll III HIM lllll lllll llll llll 61 SK98K0342 THE STUDY OF THE UPTAKE OF PLUTONIUM BY SLOVAK ZEOLITES Peter Lukáč, Niva Patzeltová1, Mária Fôldesová, Pavel Dillinger Faculty of Chemical Technology, Slovak University of Technology, Radlinského 9, 812 37 Bratislava, Slovak Republic 'institute for Preventive and Clinical Medicine, Limbová 14, 833 01 Bratislava, Slovak Republic The uptake of radionuclides from nuclear power plant station waste water is studied very intensively in recent time. Low concentration of plutonium have occured from time to time in the presence of this radionuclide indicate a damage of a fuel element. Higher concentration of plutonium would be certainly in solution in the case of some accident of nuclear reactor involving of reactor core damage. The efficiency of separation of plutonium depended on pH. There is a plenty of methods for uptake of plutonium from very strong acidic solutions. The uptake by organic sorbents in very strong acid solutions showed a very good results [1-3]. The separation of plutonium from low acidity solution (pH 5-9) has many obstacles and is not simple [4,5]. In this study were applied natural and chemicaly modified clinoptilolite and mordenite type of zeolite to sorbed plutonium from solutions of pH 5.1 - 8. All experiments were carried out with zeolitic material of the same granularity in the range of 1.2 - 2.5 mm. The sorbents were came from deposits: Nižný Hrabovec - clinoptilolite type, Rankovce and Zamutov - mordenite type (Slovak Republic) and Lipča - clinoptilolite-mordenite type (Ukraine). All natural materials were treated by: 1, 2,4 or 6 moll"1 solution of NaOH, or 1 and 4 mol.ľ' solution of NaCl, or 1 mol.ľ1 solution of 1 NaHCO3, or 1 mol.ľ solution of Na2C03. The temperature of solutions was 80 °C and the time of treatment was 4 hours. The method of model radioactive solution was used to find distribution coefficient K The distribution coefficients Kd of all sorbents mentioned above and for all used solutions of the pH from 5.1 to 8.0 were calculated. The best distribution coefficient of some type of sorbents is very low and uptake of Pu by these sorbents is negligible as well. It is a reason why these sorbents will not be mentioned more. For example, the natural zeolite fromNiDný Hrabovec modified with lmoU'solution of NaOH has two maximums of Kd -one at 5.7 pH and secod between 7 and 8 pH. Generally.all type of 62 2 ŕ RHD Jasná pod Chopkom zeolites have two maximums of Kj, only the intensity of maximums and location are different. All modifications of zeolite from deposit NiDný Hrabovec show the best value of Kd and the best sorption ability. For the real utilisation of natural and modified zeolite for sorption of Pu is inevitable to verify the sorption ability in presence of other radionuclides and chemical compounds which can be in waste water of nuclear power station. These radionuclides and chemical compounds can heavily influenced sorption of Pu from waste waters. REFERENCES 1. Yamamoto M., J.Radioanal.Nucl.Chem.2,401-408(1985) 2. Joshi S.R., J.Radioanal.Nucl.Chem.l02,190-193(1986) 3. Mátel Ľ., Mikulaj V., Rajec P., J.Radioanal.Nucl.Chem.l75,41-46(1993) 4. Barney G.S., Blackman A.E., Lueck K.J., Green J.W., Report WHC-SA- 0533,USA, 1989 5. Franta P., Kuča L., Svoboda K., Šebesta F., Nucleon 2,19-23(1996) 2 ŕ RHD Jasná pod Chopkom SK98K0343 63 THE DEVELOPMENT OF NATIONAL REGISTRATION SYSTEMS OF RADIATION PROTECTION IN THE CZECH REPUBLIC Karla Petrová, Zdenek Prouza, State Office for Nuclear Safety,Senovážné nám.9,100 00 Praha 1 Introduction In accordance with the demands of the new Czech legislation in radiation protection, which is effective from 1997 year, State Office for Nuclear Safety (SONS) develops the central registration systems of occupational radiation exposure, sources of ionising radiation and licensees. Simultaneously the system enabling the evaluation of medical exposures in the Czech Republic is created. The methods and partial results are described. Methods, analysis, results Since 1994 the Central Register of Occupational Exposure (CROE) is created [1].CROE enable us to evaluate individual as well as collective doses of radiation workers and to estimate their distribution and time trends. Workers are categorised by their profession and dose distributions are evaluated (Fig. 1). Fig.l Th» dtetrlbuilon of worktn. colltdlv* in potet QdefBCtotcopy,w«t-bgg 0,636 1,36 2.14 gw orks w kh radlolsotopsa 0,901 1,18 1.31 QNFP 1,949 1.26 0.66 Qht*Khear» 6,679 »,5o 1,44 ••p*chRted w orkt 0,607 0,«8 1.43 B not categories 0,624 0,»4 1.35 The comparison of dose distribution in 1996 and 1997 years shows the stable trend of radiation doses distribution [ Fig.2]. 21st RHD Jasná pod Chopkom We can observe, that about 30% of HE lies below MDL (0.05 mSv in the case of film dosimetry) in 1997 year [Fig.3]. Fig.3 The most frequent group of workers is created by the medical workers with the average effective dose equivalent 0.85 mSv ( however 1.44 mSv for workers with HE > MDL and for example 4.33 mSv only for cardiologists) [Fig.4]. The most cases of overexposed workers can be observed also in the group of medical workers. During the 1997 year there was registered 29 workers with HE > 20 mSv, 24 of them working in medicine. 2ŕ RHD Jasná pod Chopkom 65 Fig.4 10 , * s' • • í 's - V ; 1 . * i 0,1 - poiel97 pcOl 96 SE97 SEW HF97 t-EOC m • Its 7,423 9,205 6,20 5.95 0,85 0,73 n nm 0,829 0,907 1,34 1.51 1,62 1,66 EJrtp 0,776 0,74í 1,02 1.10 1,32 1,47 Based on the development of CROE, the registers of sources of ionising radiation and licensees are created in SONS. Data will be connected and the system of databases will provide the regulatory authority with set of information about the individual licensee [see Scheme 1]. Scheme 1: I l"^«to» j j—Li—I Praoovnt místo I CintráM agenda pncovMr '|R«gKtrei|»ilc The attention of regulatory authorities in the recent years is focused also to the evaluation of medical exposure. SONS co-operate on this field with the General Health Insurance company, which provide it with very valuable data, which enable us to sort examined persons going through the individual procedure by the age and sex [Fig.5,6]. 66 2 ľ' RHD Jasná pod Chopkom Fig.5 The work concerned to the evaluation of medical exposure is sponsored by the Grant Agency of the Czech Republic. Based on this frequency study the evaluation of effective dose to the population arising from diagnostic procedures can be estimated. This was already done for nuclear medicine procedures ( see paper Petrová,Hušák). The evaluation of doses for individual X-ray procedures is recently maintained. Fig.6 Literature l.Petrová K., Prouza Z.; The National Central Registries of Occupational and Medical Exposure in the Czech Republic, IRPA 9 Conference Proceedings, Vienna, Austria, vol.4,682-684 pp (1996) 2ľ> RHD Jasná pod Chopkom ' SK98l<0344 EVALUATION OF IRRADIATION EV PATIENT'S ENVIRONMENT RECEIVING 131I THERAPY Jozef Husár1, Ing. Alžbeta Fúriová2, MUDr. Františka Borovicová1 'State Health Institute of SR Capital Bratislava, Ružinovská 8, 820 09 Bratislava 2St. Elis. Oncological Institute, Heydukova 10, 812 50 Bratislava The radioactive 131I is used in thyroid gland treatment for cancer and thyrotoxicosis. Administered activities of the radioactive Iodine are in the former case GBq of order, in the next one of tenths of GBq. Discharging of the patient from a hospital after the treatment is influenced by the clinical parameters as well as by the residual activity 131I in his body. This residual activity must be so low that the patient does not cause the irradiating of the other people in his surrounding with effective dose exceeded 1 mSv in one year. This value is recommended by BSS and it is accepted to Slovak Republic legislation by law number 290/96. The given limit will not be exceeded if the residual activity I31I in patient's body is not over 40 MBq. At the present is abroad and also in Slovakia considered an increase of the residual activity for discharging patient from the hospital and even about a possibility of the ambulant thyrotoxicosis treatment. In this case the patient would have to keep some restrictions in behaviour to his surrounding for some time. The limits of rest activities for discharge differ in various countries from 75 MBq to 1 100 MBq. The period of necessity to keep the restrictions is also various and can take 1-4 weeks. It depends on the clinical condition of the patient, treated disease, the rest activity of 13II, the social background and the way of life in the given country. The further criteria for determination of limits and periods of restrictions and also the reason of their different values are various ways of treatment and from that resulted various administered activities radioactive Iodine which can be administered in two ways - by a single dose or it can be divided to several fractions. It's a topic of study of publications related to this theme. Similar conclusions are given in the publication of the expert group 31 of Euratom „Radiation protection following Iodine - 131 therapy" [1] and summary publications from Germany [2] and England [3]. In the publication of Euratom is suggested to accept the limit of effective dose 1 mSv, in accordance with BSS, for adults in the patient's household. For the children up to 2 years and pregnant women in the patient's household the suggested limit is 0,3 mSv. The same limit, 0,3 mSv, is suggested also for people out of patient's household. Every country or clinic should ensure not to exceed the suggested effective dose limits for persons round the patient after discharging him from the bed station. The residual activity and following restrictions for the patient's behaviour should be estimated in according to the above mentioned criteria. The next part of this article describes measurements made in the bed station of Clinic of Nuclear Medicine in St. Elis. Oncological Institute in Bratislava. Besides the other diseases, there are treated thyroid cancer and thyrotoxicosis with the use of I3II. The aim of the measurements was to determine the possibility of the ambulant treatment of thyrotoxicosis or the possibility of shortening of the patient's stay in the bed station that the effective dose would not be exceeded suggestions according to the publication of Euratom [1]. The measurements were made also with thyroid cancer patients but 68 21st RHD Jasná pod Chopkom owing to clinical reasons the ambulant treating in this case is not permissible. Therefore this article does not describe the results of these measurements. The effective dose rates were measured in 0,25 m; 0,5 m; 1 m and 2 m distances from the patient's thyroid so the effective dose in the patient's surroundings could be determined. During the measurements the patient was standing and the measuring instruments were in the high of his thyroid gland. The measurements were provided simultaneously by 2 metrologically verified dose meters: VICTOREEN 450 (300 cm3 pressurised ionisation chamber) and FAG FN 40 F2 (3 cm3 ionisation chamber). Values of the effective dose rates measured with both dose meters were averaged (average background 0,2 uSv/h was subtracted from measured value with FAG, dose meter VICTOREEN has background compensation). It was measured once a day including, if necessary, some weekends. Following parameters were recorded for every patient: age, accumulation, administered activity of I, the day administered fraction, weight of the thyroid, planimetric determined area of the thyroid. The example of the results of one patient's measurement is shown in Graph 1. The arrows mark the day of the administered fraction of l31I. In the day 0 was administered the last fraction. Patient n. 61, Thyrotoxicosis Results of measurement 66 years, 370 MBq Accutn 47%/24h T{ef) 6,4 days Thyroid 44 g 25 sq cm a VC 0,25 • ve 0,5 VC1 • VC2m o FAG 0,25 • FAG 0,5 FAG1 • FAG 2 m -Aver. 0,25 -Aver. 0,5 10 15 -Aver. 1 Time [day] -Aver. 2 m Evaluation of the measured effective dose rates: Averaged values measured from the 3rd day after the last fraction administered I3II (physiologically distributed Iodine in the patient's body) were mathematically analysed (Spreadsheet Excel 5.0). Outcoming exponential equations of the effective dose rate decay for every measured distance were used to calculate effective half- life, as well as the effective dose rates values in the half day intervals to the day of its falling below 0,5 uSv/h. These calculated values were successively summed from the lowest effective dose rate to value of 21st day, 20th day, 19th day etc. to 1st day of 131I administration. The values between the 1st day of administration and the 2nd day after the last administration of 131I is not possible to calculate according to the mathematical analysis. The absent effective dose rates were completed by measured or interpolated values. This way calculated the sums of effective dose rates could be used for the determination of the effective dose for the other person with the condition that spent time in distances 0,25 m; 0,5 m; 1 m and 2 m from the patient would be known. But it is not possible to evaluate it exactly because of various behaviour patients to other people. That is why some models of behaviour of a person to the patient were created in 2ľ RHD Jasná pod Chopkom 69 range from extreme long time in the short distance to extreme short time in the long distance. Between these extremes are models of behaviour which can be taken into consideration in real conditions. Following models were chosen for evaluation: A - 0,25 m: 4 h; 0,5 m: 8 h; 1 m: 8 h; 2 m: 4 h B - 0,25 m: 2 h; 0,5 m: 6 h; 1 m: 8 h; 2 m: 8 h C - 0,25 m: 1 h; 0,5 m: 4 h; 1 m: 6 h; 2 m: 10 h D - 0,25 m: 0,5 h; 0,5 m: 2 h; lm:10h; 2 m: 10 h E - 0,25 m: 0,5 h; 0,5 m: 1 h; 1 m: 8 h; 2 m: 10 h F - 0,25 m: 0,5 h; 0,5 m: 1,5 h; 1 m: 8 h; 2 m: 5h G- 0,25 m: 0,5 h; 0,5 m: 2 h; 1 m: 5 h; 2 m: lOh H - 0,25 m: 0,25 h; 0,5 m: 1 h; 1 m: 8 h; 2 m: 12 h I - 0,25 m: 0,25 h; 0,5 m: 0,5 h; 1 m: 6 h; 2 m: lOh J - 0,25 m: 0,25 h; 0,5 m: 0,5 h; 1 m: 4 h; 2 m: lOh K - 0,25 m: 0,1 h; 0,5 m: 0,25 h; 1 m: 2 h; 2 m: 8h L - 0,25 m: 0 h; 0,5 m: 0,1 h; 1 m: 1 h; 2 m: 4 h Each of models from A to L suppose the maximum spent time of the other person in given distance from the patient for 1 day. According to these models is possible to calculate the effective doses for other person in wide range of patient's behaviour alternatives. Besides, it is easier to explain the patient safe behaviour to other persons in this way than to some extend blurred (and not always fully named) restrictions stated in the publications of other authors. For every model of behaviour the sum of the effective dose rates for given day was multiplied by half value time in given distance (the effective dose rates summed in intervals of half a day) and the effective doses for 4 measured distances were summed. In this way were calculated effective doses for a person in neighbour of the patient according to the given model from the 1st day of administration I31 to the 21st day after the last fraction of the radioactive Iodine. The calculated values are presented in Graph 2. 66 years 1'atlentn. 61, Thyrotoxicosis Effective doses for other idivlduals 370 NBq •WYk 47 %/24h í- \ V \ T(ef)6,4day h, S s, , \ s - - -B , "S K s s ... -D • - ' t. Ss Effectiv e dos [#Sv ] mm •h - - -F m N *i • * Ls, IN m s S! ta •- 5- •- Ml ... -H s; taj •-w •i k, ta ta, ta m •m • Srn •M ř r : m Ě i - - -J -15 -10 -5 0 5 10 15 20 Hrne [day] ... -L The obtained data are the basis to determine the day of discharge the patient from the bed station related to the last day of radioactive Iodine administration. It was made for effective dose limits 0,25 mSv; 0,5 mSv and 1 mSv and for all models of the 70 2 ŕ RHD Jasná pod Chopkom patient's behaviour to the other persons. The example of an evaluation for patient number 61 is presented in Tab. 1. Eff. dose [mSv A B C D E F G H I J K L 0,25 27 23 19 15 13 13 13 12 9 7 2 0 0,5 20 16 11 8 6 6 6 5 2 0 0 0 1 13 8 4 1 0 0 0 0 0 0 0 0 To the present time the results of the effective dose rates measurements for 17 patients were evaluated by described way. The age of the patients was from 41 to 82 years, the administered quantity of 13II was from 259 to 481 MBq, in fractions 37 MBq, 74 MBq, or 111 MBq. The calculated effective half - life of I3!I excretion from the patient's body which is crucial for the length of patient's necessary staying in the bed station, were from 4,2 days to 8 days. This great extend of values is given by the different clinical parameters of the treated patients. After the analyse of them can be said that the effective half- life increases, when the patient is elder, has greater mass of thyroid and the accumulation is higher. The result of measurements is the suggestion to determine the conditions of discharging the patient from the bed station with regards to radiation safety of population. Beforehand it must be said that the suggestion is designed only to thyrotoxicosis treatment used in the therapy department of Clinic nuclear medicine St. Elis. Oncological Institute in Bratislava. For common use it is necessary to confirm it with further measurements. Introduction to praxis must be approved by competent authorities. The basis for that suggestion is considered guidance of the expert group of Euratom [1]. Further, all the patients were interviewed about their living conditions and necessary communication with people in their household and neighbourhood. It was found out that in majority of households live children and it is hard problem to ensure suitable restriction conditions for the contact of the patient with other members of the family. As the basis of these facts it is recommended discharging the patient according to the model of behaviour B and effective dose limit 0,25 mSv for the households, where children up to 2 years and/or pregnant women live. For the other households it is the model D and the effective dose limit 0,5 mSv. For these households where are no children and pregnant women and no contact with them in the period of one month is quaranted, it is possible to use model H with the effective dose limit 0,5 mSv (but interviews with the patients showed that this case is a very little probable). The major criterium for discharging the patient from the bed station is the value of the effective dose rate. Table 2 summarises data of the measured patients, the day of discharging the patient from the bed station for the model of behaviour D and the effective dose limit 0,5 mSv. There are values of the effective dose rates in 1 m distance for discharging (day 0) and the effective dose rates for days before discharging. Conclusion: At the present time we don't suggest using the ambulant treatment of thyrotoxicosis by 13II. For discharging the patient from the hospital we suggest to think criteria according to the model of behaviour D with the effective dose limit 0,5 mSv. For the households with children up to 2 years and/or pregnant women according to the model B with effective dose limit 0,25 mSv. st 21 RHD Jasná pod Chopkom 71 References: [1] Euratom: Radiation Protection following Iodine-131 therapy, Expert group 31, Euratom 1998. [2] H Schicha, Ch. Reiners: Ambulante versus stationäre Radioiodtherapie; Nuklearmedizin 1998; 37, (4). [3] L. K. Harding: Radiation protection legislation; Eur. J. Nucl. Med. 1998; 25, (2). Thyrotoxicosis I According 0 - limit 0,5 mSv, 1 m D - 0,25 m: 0,5 h, 0,5 m: 2 h, 1 m: 10 h, 2 m: 10 h Pat. n. 38 43 18 59 19 61 4 28 5 51 39 3 29 37 20 36 62 Age 80 82 42 67 50 66 65 76 53 49 59 51 61 47 52 51 41 Tef (day 8 7,7 7 6,8 6,1 6,4 6,5 7,5 6,5 5,3 6 6,3 4,4 4,4 5,7 5,2 4,2 Administered activity 1-131, mCi, resp. MBq mCi 13 11 9 10 5 10 7 11 9 11 11 9 9 10 9 10 10 MBq 481 407 333 370 185 370 259 407 333 407 407 333 333 370 333 370 370 Accumu ation %/24h, bold0/./48h PB2T76 : 67 30 66 47 _43 40 49 26 68 48 26 42 38 51 42 Thyroid area, sq cm 116| 52 45 91 24 25 44 45 32 37 44 34 28 29 Thyroid mass, g | 300| 124 98 210 40 44 95 100 58 75 95 65 50 52 Discharging day after last fraction admistered activity 19 19 17 15 10 8 8 7 7 7 6 5 5 3 2 1 0 The day before discharging - effective dose rate, |iSv/h 0 1,82 1,74 1,91 1,91 2,11 1,94 2,16 1,68 1,92 2,51 2,36 2,13 2,65 2,75 2,90 2,50 2,05 1 1,99 1,91 2,11 2,12 2,37 2,16 2,40 1,85 2,14 2,87 2,65 2,38 3,10 3,43 3,80 5,48 2 2,17 2,09 2,33 2,35 2,65 2,41 2,68 2,02 2,38 3,27 2,98 2,66 3,62 4,15 4,70 3 2,37 2,28 2,58 2,60 2,97 2,69 2,98 2,22 2,64 3,73 3,34 4,10 2,60 6,03 4 2,68 2,50 2,85 2,87 3,33 3,00 3,32 2,44 2,94 4,25 4,13 4,40 3,43 5 2,82 2,73 3,14 3,18 3,73 3,34 3,69 3,40 3,70 4,67 4,33 4,75 5,10 6 3,07 2,99 3,47 3,52 4,18 3,43 4,30 3,78 4,55 5,13 6,05 7 3,35 3,27 3,84 3,90 4fW 4 37 4 7(1 4Sfi 5S5 7,01 8 3,66 3,58 4,24 4,31 5,57 5,51 5,10 9 3,99 3,92 4,68 4,78 6,43 10 4,35 4,29 5,17 5,29 7,28 11 4,75 4,70 5,72 5,85 12 5,18 5,14 6,32 6,48 13 5,65 5,62 6,98 6,88 14 6,17 6,15 7,71 7,48 72 21" RHD Jasná pod Chopkom SK98K0345 RADIATION LOAD OF POPULATION DUE TO TREATMENT OF CHILDREN NEUROBLASTOMA BY 13II-MIBG J. Heřmamká1}, J. Zimákl), P. Kavan 2), P. Došel3) I} Clinic of Nuclear Medicine, Faculty Hospital Motol, V úvalu 84, 150 06 Prague 5 2> Clinic of Children Oncology, Faculty Hospital Motol, V úvalu 84, 150 06, Prague 5 3> Institute of Aviation Medicine - Prague, gen. Píky 1, 160 60, Prague 6 INTRODUCTION Clinic of Children Oncology, Faculty Hospital Motol, deals for several years with the therapy of children neuroblastoma using various methods like chemotherapy and radiotherapy. Since 1997, the method exploiting 131I-MIBG in the combination with the oxygenotherapy has been introduced. 131I-MIBG is administered at the Clinic of Nuclear Medicine. Its expertise in application of unsealed radioactive sources is exploited in this way. Unlike in usual treatment procedures applied at the Clinic of Nuclear Medicine, I3II-MIBG therapy involves members of common population. According to the Decree 184/1997 of the State Office for Nuclear Safety [1] they can be divided into two groups: • Family members who take care of children directly at the Clinic; for them, the annual effective dose of 5 mSv must not be exceeded. • Ambulance drivers transporting children from Faculty Hospital Motol to Institute of Aviation Medicine located in Praha-Staešovice where the hyperbaric chamber used for oxygenotherapy resides. For them as well as for the Institute staff, the annual effective dose of 1 mSv must not be exceeded. This paper describes radiation protection aspects of 131I-MIBG treatment. It presents and discusses dose equivalents found for the population groups mentioned above. PATIENTS Up to now, 5 children (3 girls, 2 boys) in the age range from 5 to 14 years underwent the treatment. To two of them, I31I -MIBG was administered twice. In one case, the time interval between two administration was 13 months, in the second case 8 months. It means that our results are based on experience gained from seven cases. TREATMENT PROCEDURE For the treatment, experience of Department of Paediatric Oncology, Emma Kinder Harhuis Hospital, Amsterdam, The Netherlands are folly exploited. The activity of 131I-MIBG from 3.7 to 5.5 GBq is administered per infusion at the Clinic of Nuclear Medicine. The 2nd, 3rd, 4th and 5th days after the administration, the child undergoes the oxygenotherapy in the hyperbaric chamber at Institute of Aviation Medicine - Prague. The commonly used therapy schedule is applied in all cases [2]. MONITORING TECHNIQUES Personal digital dosemeters STEPHEN 6000 were used for individual monitoring. Surface activity in the hyperbaric chamber and in an ambulance was measured by the apparatus CONTAMAT FHT 111 M calibrated in Bq. cm'2. 21" RHD Jasná pod Chopkom 73 RADIATION PROTECTION ASPECTS Two groups of population are involved in the treatment procedure: • family members taking care of the child treated, • ambulance drivers transporting children to oxygenotherapy and the staff of the Institute of Aviation Medicine. Family members are classified as persons who knowingly and willingly out of the framework of professional duties take care of patients treated by ionizing radiation. Their annual effective dose must not exceed 5 mSv [1]. The annual effective dose for the second group representing common population must not exceed 1 mSv. Conservative estimates of dose equivalents based on theoretical as well as experimental bases indicated that the values well below these limits have been expected. Nevertheless, we decided to check the true values through individual monitoring of both groups of population. Our decision has been justified by the lack of our personal experience with this type of treatment. Moreover, the behaviour neither children nor family members can be predicted in detail. Also, exceptional situations during the transportation and oxygenotherapy cannot be excluded. To minimize mutual personal contact of a family member and a child, the special room is reserved for family members. It is connected with the room of the treated child by an audio-visual circuit enabling two-ways communication. The apparatus in the room of a child can also be used as TV set. Whenever possible, family members stay in shifts at the Clinic. It contributes to the decrease of individual dose equivalents. A family member stays at the Clinic during the day only. At nights, the behaviour of the child is checked by nursing staff with the help of the monitor placed in the nursing room. Any contacts of the family members with other patients is strongly forbidden. RESULTS AND DISCUSSION Family members Results of the monitoring of family members are summarized in Table 1. A, L and H denote the administered activity, the lenght of the treatment at our Clinic and dose equivalent, respectively. From the Table it follows that - with single exception - dose equivalents from the external irradiation had values even below 1 mSv. The available (limited) amount of data indicate lack of correlation between the administered activity, the lenght of treatment in the Clinic and dose equivalents H. Values of H depend mostly on demands of the child for the direct contact. Measurement of 13II in thyroid glands of family members indicates no contribution of internal irradiation. Drivers With one exception, the single ambulance driver ensured the transportation of the child to and from oxygenotherapy. The time needed for single trip is of about 10 minutes. The minimum H per transportation (two ways) was 0.1 nSv, the maximum 11.4 uSv. The total value in the year 1997 was 20.8 nSv, in the year 1998 8.4 uSv. In the exceptional case, the transportation was ensured by changing drivers. The minimum H per transportation was 0.1 ^Sv, the maximum 0.3 ^Sv. From the results it follows that the total dose equivalent due to transportation represents approximately 2% of the 74 21" RHD Jasná pod Chopkom annual limit. The maximum H of 11.4 |ASV was registered when the driver helped with the manipulation with the child. For practical reasons, swapping of drivers is planned. Thus, no significant dose equivalents are expected under normal conditions. Table 1; Dose equivalents for family members Patient Year Age A L Family H No. of [y] [GBq] [d] member [mSv] 1 1997 5 5.5 8 father 0.40 mother 0.16 uncle 0.15 1998 6 3.7 9 father 0.13 mother 0.20 2 1997* 7 5.5 26 father 0.44 mother 0.47 1998 8 3.7 18 father 0.22 mother 1.44 3 1997 8 3.7 14 father 0.51 mother 0.56 4 1997 14 5.5 16 mother 0.31 5 1997 5 5.5 16 father 0.64 mother 0.20 •without oxygenotherapy Staff maintaining the hyperbaric chamber Dose equivalents for the staff maintaining the hyperbaric chamber (7 persons) ranged from 0.3 ^Sv to 4.9 jiSv in the year 1997. In the year 1998, H reached the values from 0.2 uSv to 1.9 uSv. The values depend on: - the lenght of the monitoring time which was approximately from 30 to 90 minutes, - the number of therapies at which the member of the staff was present. Results of monitoring are comparable with those obtained outside the chamber without the presence of the treated child. The maximum value 4.9 uSv was reached in the case when the member of the staff once entered the chamber in the presence of the child for operational reasons (H of 4.1 uSv was measured in this case). Hyperbaric chamber and ambulance No values of surface activity exceeding 3 Bq.cm'2 [1] were found. It could be expected because no situation leading to increased values (e.g. vomiting out of the plastic bag, etc.) occurred. CONCLUSIONS The presented results indicate that no danger of exceeding limits of dose equivalents in the sense of (1) occurs under normal conditions. The shortest recommended time interval between two therapies is 6 months. According to the results obtained up to now, dose equivalents well below the limits are expected even in the case of two therapies. It could seem that the continuation of our work is more or less superfluous. We have decided to continue in non-standard monitoring because of the following reasons: 2ľRHD Jasná pod Chopkom 75 • we have had restricted amount of data at disposal so that our conclusions are of preliminary nature only, • we have no experience with children younger than 5 years - the situation might be different in such cases, • the occurrence of accidents cannot be excluded and we have to learn their danger. REFERENCES 1. Official Journal of the Czech Republic No. 66: Degree 184 of the State Office for Nuclear Safety on Requirements on Security of Radiation Protection, 1997 (in Czech). 2. Došel P.: Personal Communication. Prague, 1998 ACKNOWLEDGEMENT This research has been partially supported by Granting Agency of the Ministry of Health of the Czech Republic, grant No. 3943-3. 76 SK98K0346 2ľ"RHD Jasná pod Chopkom ABOUT CALCULATION OF THE EQUIVALENT DOSE FOR AUGER EMITTERS USED IN DIAGNOSTIC NUCLEAR MEDICINE RNDr. Soňa Ptáčníkova, CSc, Ústav preventívnej a klinickej medicíny, Bratislava Introduction Nuclear medicine procedures provide valuable diagnostic information, but as with any medical procedure, the risk and benefits must be weighted. The values of effective, equivalent and absorbed doses are the essential for assessing the risk involved in using these procedures. Currently used MRD ( Medical Internal Radiation Dose) formulation assumes a homogenous distribution of radionuclides over the human organs and calculate the average radiation dose to the organs and tissues (what means that all cells in organs are assumed to receive the same dose). The reality is somehow different. At the tissue level, it has long been recognized that many common diagnostic agents are not distributed uniformly in organs and tissues. Indeed, diagnostic agents have been designed to target specific tissues within the organ (precapillary arterioles, phagocytic elements, renal cortical tubules). That means when calculating patient risk during diagnostic procedures, the parameter that may be of particular significance is the dose to the radiosensitive cells ( the steam cells) of the bone marrow. In case, when the Auger emitters are involved, it is important to take into account not only distribution on organ, multicellular, cellular or subcellular level , but we need to define more accurately the location and spatial dimension of the primary radiosensitive targets in the cell to allow a direct correlation between the absorbed dose from Auger emitters and the biological effect. These considerations led to the proposal that the mean organ absorbed dose for Auger emitters be corrected with a factor which depends on the fraction of activity in the organ which is bound to DNA. It is therefore necessary to developed the risk assessment methods and recalculate the values of equivalent dose for in diagnostic nuclear medicine involved Auger electron emitters. Dose nonuniformity The tissue distribution of radionuclides depends on used radiopharmaceuticals and for some of them it is not homogenous at all. This doesn't create a problem as far as we have radionuclides emitting the photons of sufficiently high energy (when the range of photons is much longer than cell diameter) and the value of mean absorbed dose is appropriate for estimation of risk associated with diagnostic procedure. The MIRD dose calculation can be inadequate for particular radiation and photon of low energy. In these cases the individual cell absorbed doses differ significantly from calculated average doses. Recent experimental data and dosimetric calculations for several radiopharmaceuticals have indicated the great variability in the dose to individual cells within the organ (4), which depends on size and affinity of the carrier molecule to the certain cellular components, on the fraction of tissue volume occupied by the radiopharmaceutical, on type of the emitted particles during the radionuclide decay. For example, the radiation 2ŕ RHD Jasná pod Chopkom 77 dose to the cells that concentrate20 1 Tl is substantially higher than the MIRD mean dose (5). The similar situation is in case of using the Tc labeled albumin colloid for examination of liver, lung and spleen. The macrophages, which are the main site of 99mTc uptake (studied by microautography) obtained the absorbed dose 10-60 times the MIRD estimates (4). In case of the calculation for "m Tc labeled microspheres and macroaggregated albumin found that the doses to individual lung cells varied substantially form the mean dose estimate to lung. 92 % were shown to receive a dose approximately one-forth that assumed the MIRD calculation, while remaining 8 % received the doses from 3 to 7,500 times the mean dose, amounting to hundreds and thousands of cGy in some instances (4). Such findings make it important to understand the risk of mutagenesis at the cellular level as a function of absorbed dose.What can be the implications for radiation protection ? Calculation of effective dose for Auger emitters The problem is even more dramatic. The absorbed dose calculated at the cellular or organ level can not be used alone to predict the complex variety of biological responses caused by Auger emitters. The location and spatial dimensions of the primary radiosensitive targets in cells must be established more accurately and also the location of the radionuclide relative to the target identified. This considerations led to proposal that the mean organ absorbed dose for Auger electron emitters be corrected with the factor that depends on the fraction of DNA bound activity. The American Association of the Physicists in Medicine (AAPM) in their report (3) adopted the method for calculation of equivalent dose for Auger emitters and also reviewed the effects of Auger electrons in several biological systems and recommended radiation weighting factor of 20 for stochastic effects. As we mentioned in our previous paper (6) the expression for equivalent dose for Auger electron emitters was derived HT - HTjiAnger + Hx,Rotlier- [ l+í> (WRAuger-1)] DT)R Aoger + D T, R other Where ft is the fraction of organ activity bound to DNA, wR Auger is the radiation weighting factor for Auger electrons, H T,R Auger and DT,R Auger are the equivalent dose and absorbed dose from Auger electrons and HT,R other and D T, R other is the equivalent dose and absorbed dose from others radiation. Here is necessary to notice that the values of absorbed dose were calculated using MIRD (Medical Internal Radiation Dose) formulation assumes a homogenous distribution of radionuclides over the human organs and calculate the average radiation dose to the organs and tissues. In this contribution we would like to demonstrate the calculation of the mean equivalent dose and compare these values to the mean absorbed doses, for the testes of male patient who is undergoing a cardiac imaging procedure with201 Tl-thallous chloride and for the liver of patient who is undergoing the liver scintigraphy using the 99mTc MicroUte. Examples MYOCARD SCINTIGRAPHY Assume a male patient undergoing the cardiac imaging procedure with intravenously injected 111 MBq of 201 Tl-thallous chloride and also assume that the non-testicular 78 2 ŕ RHD Jasná pod Chopkom activity is distributed uniformly throughout the body. Suppose that 50% of testicular activity is bounded to the DNA (9). We would like to calculate the equivalent dose to the testes using the weighting factor for Auger electrons WR = 20 and consequently compare this value with value of mean absorbed dose to be able to see the effect of underestimating the local effect of Auger emitters when incorporated in cell nucleus. Exam code radiopharmaceurical activity [MBq] MYOCARD SCINTIGRAPHY (5400) 201T1-Thallous chloride 111 WR 20 Calculated values Fo 0.5 HT^ f mSvl 132.43 A**. [Bq si LIVER SCINTIGRAPHY The second case is the liver scintigraphy using the 99mTc-S-coloid (Microlite). The mean applied activity is 200MBq 99mTc-S-coloid. From literature we assume the following activity uptake - 70% in liver, 10% in spleen, 10% in red marrow and 10 % in remaining tissues. According the paper (1) and microautoradiographic study (7), the Kupffer cells ( macrophages), one type of cell which lines the hepatic sinusoids, are thought to be the main site of 99mTc uptake in liver. According (1), the fraction of the liver volume occupied by the labeled cells was derived and found from 0.1 to 1 %. In the "worst scenario" we assume that all this activity is bound to the DNA so ft is 0.01. We calculate the "S" values for liver using the MIRDOSE 3 code. Exam code radiopharmaceutica] activity [MBq] 1 LIVER SCINTIGRAPHY (5439) 99mTc-S-coIloid 200 | WR 20 Calculated values Fo 0.01 HUver [mSvl 173 m 12 Au«r [Bq si 4.4xlO Duvr [mGyl 17.21 W u A Redmam-w [Bq Si 6J5xlO HLlv,r/DIiv„fSv/Gyl 1.005 (8) AR,„lůltaIKMn, pqs] 6.25x10" As^ fBqsl Fig.l: Dose histogram in human liver cells from 99mTc% - Microlite DOSS HISTOGRAM IN HUMAN LIUER CELLS FROM Tc-99m MICROLITE ífle • o.i i ie CELLULAR-TO-CONUENTIONAL DOSE RATIO But this doesn't mean that in the case of 99mTc is not important to take into account the problem (already mentioned in this contribution several times) connected with the MIRD assumption about the homogenous distribution of radionuclides over the human organ of interest. In case of liver scintigraphy performed with 99mTc- Microlite the distribution of radionuclides has highly inhomogeneous pattern. According the (1) fraction of the volume liver occupied by the labeled cells was found to be 0.1 to 1 %. When one calculates the doses to the individual liver cells ( cellular dosimetry) and compare with conventional calculation ( MIRD), one can see the MIRD underestimation 80 21" RHD Jasná pod Chopkom of radiation doses delivered to the labeled macrophages by factor of 10 to 30. At the same time the conventional calculation overestimates slightly the dose delivered to the unlabeled cells. Fig.l shows the histogram of fraction of human liver cells versus radiation dose that cells fraction receives according the (1). Such deviation occurs whenever the range of emitted radiation is comparable or smaller than the regional inhomogeneous deposition of radiopharmaceuticals. So "the histograms showing the fraction of cells within a specific tissue receiving a given dose would be a natural extension of the single-value S factors" (1). Conclusion I. For radionuclides that deposit their emitted energy in a specific tissue predominantly by low energy Auger and internal conversion electrons (E <10 keV), the MIRD S factors cannot in general be applied at the cellular level without producing some deviation from the prediction of cellular dosimetry. These S factors have to be replace by histograms showing the fraction of cells receiving the given radiation energy. II. In calculation of equivalent doses for Auger emitters it is necessary to take into account the AAPM recommendation, which suggests the WR value for stochastic effect of Auger electrons to be 20 and involving the value of fraction of DNA bound activity of these emitters. III. To be able to satisfy these two points it is necessary to determine the cellular and subcellular distribution of each radionuclide - pharmaceutical combination in human tissues. IV. Explore risk of mutagenesis at the cellular level as a function of cellular and subcellular absorbed dose. V. Design the epidemiological studies that would address the possible consequences of dose nonuniformity, which occurs during diagnostic nuclear medicine examinations Literature 1. Makrigiogos, G.M., Adelstein, S.J., Kassis, A. I.: Limitations of MRD at the cellular level. Proceedings of Symposium ..Dosimetry of Administered Radionuclides, Washington: American College of Nuclear Physicians, 1990,44-57. 2. Goddu, M.S., Howell, R.W., Rao, D. V.: Calculation of equivalent dose for Auger electron emitting radionuclides distributed in human organs, presented at 3-rd Symp. On Biophysical Aspects of Auger Processes, Lund, Sweden, 1995. 3. Humm., J.L., Howell, R. W., Rao, D., V.: Dosimetry of Auger electron-emitting radionuclides: Report No. 3 of AAPM Nuclear Medicine Task Group No.6., Med. Phys., 21 (12), December 1994, 1901-1915. 4. Makrigiogos, G.M., Adelstein, S.J., Kassis, A. I.: Cellular Radiation Dosimetry and Its Implications foe Estimation of Radiation Risks; Illustrative Results with Technetium 21st RHD Jasná pod Chopkom 81 99m - labeled Microspheres and Macroaggregates, JAMA, 1990, vol. 264, no.5, 592- 595. 5. Rao, D. V., Sheostone, H.B., Wilkins, B.H., Howell, R. w.: Kinetics and Dosimetry ofThalium-201 in Human testes. J.NucLMed., 1995, 36., 607-609. 6. Ftacnikova, S.: Radiopharmaceuticals in Nuclear Medicine Diagnostic Procedures and the Problem of Radiation Protection, Journal of Radioanalytical and Nuclear Chemistry, Articles, vol.209, No. 2,1996,355-360. 7. Hindie, E., Colas-Linhart, N., Petiet, A., Bok, B.: Microautoradiographic Study of Technetium-99m Colloid Uptake by the Rat Liver. J.Nucl.Med., 29;1988,1118-11121. 8. International Commission on Radiological Protection. Radiation Dose to the Patient from Radiopharmaceuticals. Oxford: Pergamon Press; ICRP Publication 53, 1988. 9. Rao, D.V.,Govelitz, G.F., Sastry, K.S. R.: Radiotoxicity of thalium-201 in mouse testes: Inadequacy of conventional dosimetry. INucLMed. 1983; 24, 145-153. 82 2 ľ RHD Jasná pod Chopkom SK98K0347 INITIAL EXPERIENCE WITH SOFTWARE SYSTEM JODNEW FOR EVALUATION BIOPHYSICAL CHARACTERISTICS RELATED TO TREATMENT OF CARCINOMA OF THYROID GLAND BY I31I J. Hehnanská, J. Zimák, L. Jirsa*, M. Kárny*, K Vošmiková, J. Nimec, T. Blažek Clinic of Nuclear Medicíne, Faculty Hospital Motol, V úvalu 84, 150 06 Prague 5, CZ *Inst. of Information Theory and Automation, AVER, POB18, 182 08 Prague 8, CZ INTRODUCTION Our research tries to exploit sophisticated methods for a balancing of positive and negative consequences of radionuclide applications. We have tailored Bayesian data processing in order to support decision making during treatment of thyroid diseases with help of 131I. After successful experimental phase we have implemented them [5] . This novel in-house developed software system JODNEW is now tested. It aims at: (i) increasing quality of raw biophysical data exploited in diagnostics and therapy of thyroid diseases; (ii) estimating cumulated activity so that MIRD methodology [10] can be well used; (iii) decreasing working load on staff. JODNEW is an extensive data-base system co-operating with advanced estimation algorithms coded in C++. The Bayesian methodology adopted allows us to exploit expert knowledge, models of observed processes as well as measured data in a consistent way. This is important in the considered case when the number of measurements is quite limited and influence of biological and physical variations is high. Moreover, all estimates are qualified by the remaining uncertainty [6,7]. During diagnostics: The (functioning) volume of thyroid gland and body mass are measured. A diagnostic amount of 131I is administered. Three whole body measurements of elimination rate by urine (excretions) are made within 2 days after administration. The accumulated activities above thyroid gland and other lesions are registered within several days. Evaluation and measurements during therapy are: The accumulation ability is evaluated using diagnostic data. Consequences of' I administration are judged, then, the therapeutic activity is selected and administered. The accumulation dynamics is supervised and reaching radiohygienic limits influencing patient regime is predicted. The common features of these steps are: (i) Individual measurements are corrupted by a high and varying uncertainty, (ii) The number of measurements is limited, (iii) A significant expert experience is available, (iv) The subsequent medical decisions have to be supported also by information about uncertainty of data. APPLICATION OF BAYESIAN ESTIMATION AND PREDICTION The mentioned features are uniquely respected by Bayesian estimation [9] as it exploits consistently theory, data and expertise for good estimation; assesses uncertainties to estimates even for small data samples. The Bayesian paradigm relates the observed data D to unknown variables Q of interest by the probability density function p(D\Q) (pdf) and models vague expert knowledge about Q by the prior pdf p(Q). It uses deductive rules of probability theory for computing the posterior pdf according to the Bayes rule p(Q\D)=C(D)p(D\Q)p(Q), C(D)=a normalising constant. p(Q\D) represents the subjective knowledge corrected by the objective data. The 21" RHD Jasná pod Chopkom 83 prediction of future observations Df is given by the p(Df\D)= £Q p(Df\Q)p(Q\D). It provides medical doctors a highly expected range of an uncertain quantity through the confidence interval I=(Qe -p,Qe +p) such that p(Q flp I\D) =confidence level close to 1. The midpoint Qe of I serves as a good point estimate of the unknown Q and the half- width p as standard deviation. Summary of key achievements for individual biophysical quantities encountered is below. The common contributions of the Bayesian methodology (consistency, use of all information and information on the precision of outcomes) apply [1]. Thyroid mass [2]: Data D available are a subjective palpation, outputs from sonographic measurements and pre-processed results obtained from planar-camera images. Estimated parameter Q is the mass of the (functioning) thyroid gland. Log- normal model p(D\Q) reflects well: (i) non-negativity of the estimated mass; (ii) high uncertainty of individual measurements. Prior information modelled by self- reproducing log-normal distribution reflects experience on the mass distribution in ill population. Achievements consist of: (i) the ability to combine different techniques with a good impact on the precision of results; (ii) the enhanced search for a reliable and routinely applicable ways of measurement. Activities and excretions [1]: Data D consist of counts registered above lesion or whole body. Data from the background subtraction and calibration are needed, too. Estimated quantity Q is the instantaneous activity. Poisson modelp(D\Q) is adopted. Its mean value is the product of the unknown activity and calibration factor. Prior information is modelled by a uniform pdf. The known range is implied by the knowledge of the administered activity and radioactive-decay law. Achievements consist of: (i) a non-reduced use of the calibration information; (ii) a use of the physically justified prior knowledge. Accumulation dynamics [3,6,7]: Data D consist of a sequence of activity measurements. Estimated Q parametrises time course (t) of f(t,Q) of the unknown activity. Often, Q-(A(ti),Tef)={3sXváiy at time ti, effective half-life) but more adequate models were proposed, too [3,8]. Poisson model with the conditionally independent measurements given by the expectation/^, Q) is relevant. Prior information is modelled by a uniform pdf on a physically justified range, e.g. Tef e <0, physical half-life of the 131I>. Achievements consist of: (i) a suppression of meaningless estimates and improvement of estimation precision; (ii) a design of predictors of the moments when hygienic limits are reached; (iii) a design of better still feasible models. Compound quantities and MIRD [4,10]: Data D consist of union of all discussed measurements. Estimated parameter g is a compound quantity as specific irradiation of blood. Non-linear deterministic combinations of the previously discussed quantities define individual values of interest. For computation reasons, an engineering combination of the point estimates and their uncertainties are used. Prior information is introduced through the knowledge on constituents. Achievements consist of: (i) a development of non-trivial techniques suitable for multivariate applications of the Bayesian statistics under processing-time restrictions; (iii) the decision that the MIRD methodology [10] will be used: this will enhance the efficiency of our results and bring a new facet in the MIRD use [4]. 84 21" RHD Jasná pod Chopkom TECHNICAL DESCRIPTION OF JODNEW JODNEW implements discussed data processing. It consists of a non-standard combination of data-base language FoxPro 2.5b with estimation algorithms written in C++. The package needs at least PC 386, colour monitor, 2MB of RAM, Windows 3.xx, 10MB on hard disk = a space for data archive. The communication with user is a Borland-like style. The program starts by calling a batch file that opens the initial screen and the main menu bar with the key options: Archive Solution Application Measurement Evaluation Service. Choice is made by mouse or by hot keys. Archive serves: for input and editing of administrative information on patient stored in file 'patients' and for transferring into the files 'actual', 'record', 'measured'. The file 'actual' contains records on the treated patients and patients being prepared for the activity administration. Files 'record', 'measured' store the administrative data and results of measurement and estimation (including uncertainties). They are read-only. User's options form administrative patient data. Example: The data displayed and modified (if it makes sense) are: Full name; Evidence number (duplicity is checked); Insurance (selection from a menu); Address; Remark ; Body mass; Thyroid gland mass by palpation; Thyroid gland mass by sonography; Thyroid gland mass by camera; Diagnostics type; Administration time; Administered activity; Therapy type; Administration time; Administered activity (if available, all data are taken from file 'actual'). Solution serves: for input and editing of data related to solutions of 13II used both for administration and calibration. They are in files 'solutions', 'standpac' and 'standexc'. User's options are data on solutions of 131I and standards used for calibration. Example: The data displayed and modified are: Solution number (the latest one is offered with corresponding values; for a new one, they have to be filled); Time of solution measurement (system time is offered); Sample activity (checked range 5000 MBq); Sample volume (checked range [10,20] ml); Overall volume of solution in ml (checked range [50,200] ml); Overall activity; Specific activity (both computed in FoxPro); Current time (system time offered); Remaining overall volume; Remaining overall activity; Current specific activity (all these values provided by FoxPro). Application serves: for a completion of administrative, biophysical and medical data before and after administration of 13II. It deals with files 'actual' containing administrative data, measurement results and estimates of the available biophysical data (including uncertainties). It also updates the file 'solutions'. User's options: biophysical and medical data. Measurement serves: for recording the number of impulses registered for the patient (above lesions) and excretion. The measurements of the corresponding standards are included. These data are stored in files 'meas-lesions', 'meas-excretions'. Lesion names which are not present in a ready list are in file 'lesion'. Measurements of standards are stored in 'stan-meas-lesions' and 'stan-meas- excretions'. User's options: measurement conditions and measurement results. Example: Excretion Measurement: Administrative data (source 'patients'); Measurement number (l,2,3=daily sampling, provided by FoxPro); Measurement time (system time offered); Number of impulses registered; Decay-corrected current activity (source FoxPro); Estimates of self-calibration factor and relative excretions; Uncertainties of estimates (both evaluated by procedures in C++). Evaluation serves: (i) for evaluation of activities, medical and biophysical quantities for Daily Evaluation as well as predictions when radiohygienic limits will be reached (for all patients with 13'i administered); (ii) for Application Protocol of patients administered by diagnostic activity and for Protocol on Therapy after terminated therapy, (iii) for storing biophysical data and estimates with their uncertainties in files 'actual', 'evaluation', 2ľ RUD Jasná podChopkom 85 'meas-lesions', 'meas-excretions', 'record', 'measured' and 'limits' (xadiohygienic regime limits). User's options: the evaluation and output types. DISCUSSION AND LESSON LEARNED Here, our observations, recommendations as well as experience are briefly summarised. Data-base aspects: The GIGO principle—Garbage In Garbage Out—is the leitmotif of the measures discussed here. Quality assurance of data processing is significantly enhanced if the elementary input data are as error free as possible. Tools: (i) patient data are unambiguously identified trough a unique and checked evidence number; (ii) any piece of information is fed in once only (see transfer of information elements reflected in Archive); (iii) individual input data are checked whenever possible (see usual ranges of activities and volumes in Solution); (iv) even elementary operations and recordings are performed in a computerised way (see counting of days in Measurements of Excretions); (v) the built-in back up of data is complemented by a rescue tools available under Service option. Algorithmic aspects: Data processing must not be an additional source of errors. Tools: (i) internal consistency of the Bayesian treatment is of importance in this respect; (ii) programming emphases numerical robustness of the involved algorithms that have to be prepared to extremely wide range of input data. This motivates use of C++ and careful inspection of any simplification done. Communication aspects: Human factor is decisive for a reliable use of any software and this fact has to be respected in design. Tools: (i) Czech language is the must for all communication; (ii) warnings caused by check of input data have to give a clear guide- line what should be done; (iii) the communication has to be simple and resemble some wide spread tools; (iv) various interaction ways have to be opened; (v) minimum amount of abbreviations and familiar titles have to be used; (vi) prepared menus (see insurance in Archive) should be used whenever possible; (vii) unusual outcome should be made similar to known one (the confidence interval / is expressed as the point estimate Qe and standard-like deviation p); (viii) a long testing period is needed and any comment of the maintaining staff has to be considered seriously. Development aspects: Development time of JODNEWhas been much underestimated as it often happens with such projects. All available measures have to be employed in order to reach a success. Tools: (i) theory to be implemented has to be well structured with clearly specified case- dependent options (this is the case of Bayesian estimation); (ii) medical practice to be supported has to be well stabilised (treatment of thyroid gland diseases fulfils this requirement); (iii) the addressed set of problems and a co-operation with the software tools in use have to be carefully considered beforehand (our weaker point); (iv) the chosen construction has to be flexible as changes of such a system are more rule than exception; (v) raw data have to be produced as by-product of the system use as they are needed in further development [1]; (vi) a co-ordination of the needed experts from various branches (medical doctors, physicist, mathematicians, programmers, technologists) is the decisive and the most tough problem to be solved: it is unfeasible in routine conditions unless the developing team take the job as an interesting personal challenge; (vii) a support of a professional software house is needed at a proper time (not too early when the problem is vague and not too late when the non-expert solution costs much). Lesson learned: In spite of the problems mentioned, the positive contributions dominate. The development of JODNEW: (i) improved precision and reliability of all steps of data collection and evaluation; (ii) revealed weak points of the measurement process and led to definite improvements [2]; (iii) designed evaluations of 86 2ľ' RHD Jasná pod Chopkom the individual biophysical quantities and gave better results [7]; (iv) stimulated design of novel models of accumulation dynamics [3,6,7,8]; (v) provided an important non- trivial case study of applied Bayesian statistics; (vi) shown the way how to combine Bayesian methodology with the advanced MRD methodology [10]: both point estimates and their uncertainties can be fed in the corresponding software giving MRD results with their uncertainties included; (vii) opened a way for a systematic study of influence of the dosage of !31I for therapeutic purposes. References: [1] J. Heřmanská: Bayesian Approach to Dosimetric Data Evaluation for Medical Use of 131I. Clinic of Nuclear Medicine, 2nd Medical Faculty, Charles University, Prague, 1993. Associated Professor Thesis, 103 pp. In Czech. [2] J. Heřmanská, T. Blažek, M. Kárny, P. Vlček, J. Němec, V. Ullmann, H. Křížová, K. Vošmiková: Problems of thyroid mass estimation. In: Abstracts of the XXXIHrd Days of Nuclear Medicine, p. 2. Hradec Králové, 1996. [3] J. Heřmanská, M. Kárný, T. V. Guy, L. Jirsa, T. Blažek, J. Nimec: Biokinetics of 131I in human organism. Journal of Radioanalytical and Nuclear Chemistry, 209(2):347- 253, 1996. [4] J. Heřmanská, M. Kárný, K. Vošmiková, L. Jirsa, M. Šámal: Biophysical inputs into the software MIRDOSE. In: Sborník lékaeský, 99,1998. to appear. [5] J. Heřmanská, T. Blažek, H. Havlasová, L. Jirsa, M. Kárný , J. Zimák: System JODNEW- a guide. Technical report, Prague, 1996.128 pp. In Czech. [6] J. Heřmanská, M. Kárný: Bayesian estimation of effective half-life in dosimetric applications. Computational Statistics and Data Analysis, 24(5):467-482,1997. [7] J. Heřmanská, M. Kárný: Effective half-life estimation: Bayesian solution. Radiation Protection Dosimetry, 62(4):223-232,1995. [8] J. Heřmanská, M. Kárný , T. Blažek, J. Nimec: Progress in modelling of biokinetics of 13II. In: H. Bergmann, A. Kroiss, H. Sinzinger, editors, Radioactive Isotopes in Clinical Medicine and Research XXII, p. 367-370. Birkhauser Verlag, Basel, 1996. [9] V. Peterka: Bayesian system identification. In: P. Eykhoff, editor, Trends and Progress in System Identification, p. 239-304. Pergamon Press, Oxford, 1981. [10] M.G. Stabin: MIRDOSE: Personal computer software for internal dose assessment in nuclear medicine. Journal of Nuclear Medicine, 37(3):538-546,1996. Acknowledgement: This research has been supported by Granting Agency of the Ministry of Health of the Czech Republic, grant No. 4581-3 and by AV CR grant No. Kl 075601. *i 1SÍ nrm T ' j/-IT i umím nm •!• mil limn in ••i»"ii"i ••••'••• •••• n>7 21 RHD Jasna pod Chopkom cíKQRK0'?48 3 D GEL DOSIMETRY-METHOD REVIEW AND OUR FIRST EXPERIENCE Josef Novotný Jr'., Václav Spěváčefr, JosefNovotný*"', Tomáš Čecháfŕ, Josef Vymazať, Jaroslav Tintěraá "Hospital Na Homolce, Stereotactic & radiation neurosurgery department, Roentgenova 2 PRAGUE 5150 30, CZECH REPUBLIC Technical University of Prague, Faculty of nuclear sciences & physical engineering, Department of dosimetry & application of ionising radiation, Brehová 1 PRAGUE 1 110 00, CZECH REPUBLIC 'Faculty Hospital Motol, Radiotherapy department, V Úvalu 84 PRAGUE 5 ISO 00, CZECH REPUBLIC dInstitut for clinical & experimental medicine, Magnetic resonance unit, Vídeňská 800 PRAGUE 4140 59, CZECH REPUBLIC REVIEW OF 3D GEL DOSIMETRY Recently developed new techniques in radiotherapy e.g. 3-dimensional dose calculation algorithms, inverse therapy planning, multileaf collimators, stereotactic radiotherapy and radiosurgery are applications which demand special properties of radiation dosimeters. All the above mentioned techniques can be included under the heading of the "conformal radiotherapy". While the modern techniques of conformal radiotherapy can create rather complicated 3-dimensional dose distributions, there are not available dosimeters which can continuously measure 3-dimensional absorbed dose distributions in phantoms in arbitrary geometries and which are capable of integrating the dose. Almost every known effect of ionising radiation on matter, whether it be physical, chemical or biological, has been suggested as a basis for dosimetry, especially if the matter is (or can be made) dosimetrically tissue-equivalent. The latest candidate for our attention is the magnetic resonance effect, the dosimetric application which has been pioneered by British (Gore et al) [1], American (Appleby et al) [2] and Swedish (Olsson et al) [3] workers. The dosimeter gel and its use together with nuclear magnetic resonance (NMR) is a new and promising tool and an attempt to satisfy the requirements of the ideal dosimetry system. The method can be described as is presented in Figure 1. Unlike every other method of dosimetry, NMR gel dosimetry is totally non-invasive. There is no need to introduce a probe into the phantom, nor it is necessary to remove part of the irradiated material for testing because the phantom itself forms the detector. The atomic composition and electron density of the gel are such that it is quite close to water- equivalence. The site of measurement is determined entirely by the measuring system, which can be programmed to scan the complete 3-dimensional dose distribution. By contrast, the ionometric method, for example, requires not only the introduction of an air cavity with its enclosing walls but also a supporting structure and connecting cable. Correction for the resulting perturbations are then necessary - but are not always applied because of difficulties in estimating their magnitudes. There may also be severe difficulties in scanning an anatomically shaped phantom with an ionising chamber and it is usual to confine such measurements to a standard semi-infinite water phantom. Alternative techniques such as TLD and chemical dosimetry involve transfer of the detector from the phantom to a measuring device and it is impossible to study the 88 2ľ' RHD Jasná pod Chopkom complete pattern of dosage except by sampling on a rather coarse grid of points. The non-invasive character of NMR gel dosimetry means that it is free from these problems, but it is a relative type of dosimeter. NMR gel dosimetry shares with the photographic method the property that the complete pattern of the dose distribution - at least in two dimensions - is preserved and can therefore be sampled repeatedly and interactively, concentrating on regions of special interest. However, the introduction of a photographic film represents a considerable perturbation of the tissue-equivalent phantom and gives rise to severe problems for quantitative photographic dosimetry. By contrast it is evident that NMR gel dosimetry has a great theoretical and practical potential for the assessment of 2- dimensional and 3-dimensional dose distributions - so called "pattern dosimetry. gel dosimeter dosimeter is filled dosimeter is is prepared in the anatomic phantom as in clinical irradiated gel dosimeter NMR image of the phantom absorbed dose is analysed in NMR is obtained (Ti, T2 maps) is calculated from NMR image Figure 1 Schematic drawing of the use of gel dosimeter for absorbed dose distribution measurements. 2la RHD Jasná pod Chopkom 89 NMR gel dosimetry can be divided based, on the composition of the gel detector into two groups: Fricke-infused gel dosimeter and polymer-gel dosimeter. Fricke-infused eel dosimeter The dosimeter is formed by Fricke solution which is fixed in a tissue-equivalent medium represented by a suitable gel (agarose, gelatine). Olsson et al [3] used for ferrous sulphate solution preparation following components: -1.5 mM of ferrous ammonium sulphate, 1 mM of sodium chloride and 50 mM of sulphuric acid. Changes in the ferric ion concentration can be detected by NMR because the conversion of ferrous ion to ferric in gel after irradiation alters the NMR relaxation times. The use of Fricke-infused dosimeter gels as dosimeters in clinical applications is limited by the diffusion of ferric ions through the gel matrix, instability of the gel with time, the low sensitivity of the dosimeter system, and the variable response of the dosimeter. Polymer-eel dosimeter A new type of gel dosimeter for recording radiation dose distributions by NMR, based on a different, though related principle than Fricke-inŕused gel dosimeter, was introduced by Maryanski et al [4] in 1993. This new system is based on radiation- induced polymerisation and cross-linking of acrylic monomers which are uniformly dispersed in aqueous gel. The composition of polymer-gel dosimeter is based on acrylamide monomer (or acrylic acid) and N,N'-methylene-bisacrylamide cross-linker fixed in gelatin matrix. The formation of cross-linked polymers in the irradiated regions of the gel increases the NMR relaxation rates of neighbouring water protons. Therefore, the radiation-induced polymerisation in polymer gels plays a role similar to that of the radiation-induced oxidation of ferrous ions in Fricke-infused gels, but with four important advantages. First, whereas in Fricke-infused gel dosimeter ferric ion diffusion leads to significant blurring of the images of radiation fields within minutes after irradiation, in polymer gel dosimeter the spatial distribution of NMR relaxation rates, which reflects the distribution of dose, is stable and does not change with time. Therefore polymer gels may be imaged at any time after irradiation, which is more convenient and results in greater spatial accuracy, particularly in regions of a high dose gradient. Second, Fricke-infused gels have an intrinsically high electrical conductivity caused by the presence of ions from sulphuric acid, sodium chloride and ferrous sulphate, each at a concentration of the order of 1 mM or more. Consequently, the radio frequency field is strongly attenuated in these gels, producing significant variations of the radio frequency pulse flip angle throughout the gel volume and decreasing the signal-to-noise ratio, both of which impair the accuracy of dose measurements based on NMR relaxation data. By contrast, the polymer gel dosimeter does not contain ionic species and it shows insignificant radio frequency attenuation. Third, polymerised regions can be seen visually, as the cross-linked polymer is insoluble in water and precipitates from the aqueous phase of the transparent gel, which therefore becomes increasingly opalescent (and ultimately white) as the radiation dose increases. Therefore the gel can also be used for qualitative testing by visual inspection, and possibly, with the use of an optical densitometer, in quantitative measurements. Fourth, polymer-gel dosimeters are considerably more sensitive to radiation than Fricke-infused gels. Moreover, their sensitivity increases with the magnetic field strength, as opposed to the Fricke-infused system. Therefore, at higher imaging fields, which are increasingly common in clinical practice, the polymer gel dosimeter should be both more accurate, due to an increased signal-to-noise ratio, and more sensitive. 90 2ŕ RHD Jasná pod Chopkom However, there are some limitations as well. The first one is the temperature effect. Whereas the temperature at the time of irradiation does not appear to effect the dose response, the temperature at the time of NMR imaging has a significant effect. The second is inhibition effect that oxygen has on the free-radical polymerisation. This effect can be minimised by using deoxygenated water for gel matrix preparation and deoxygenated vessel where gel dosimeter is filled into. The vessel for gel dosimeter should be made from non permeable materials (e.g glass) for atmospheric oxygen to minimise oxygen inhibition. The third is photopolymerisation effect of the monomers. This effect can be prevented by covering the gel dosimeter vessel by a light proof cover. OUR OWN EXPERIENCE WITH 3D GEL DOSIMETRY Before starting with own 3D gel dosimetry two important quality factors of the NMR scanner had to be evaluated: spatial distortion effect and absolute and relative precision of the relaxation time measurements. The spatial distortion effect (geometric inaccuracy) of NMR image can be caused by some imperfection of NMR scanner or by investigated object itself. There are two basic potential causes of geometric distortion relating to NMR scanner: gradient field nonlinearities and resonance offsets. However, more important is the geometric distortion caused by inhomogeneity induced by the imaged object. The distortion in this case depends on both the material present in the imaged volume and the shape of the structure being imaged. In practise, gradient field nonlinearities and magnetic field inhomogeneities induced by the imaged object are the two most important sources of distortion. A special cubical perspex phantom filled by water (water is almost equivalent to the gel dosimeter composition) with the insert of array of 81 solid perspex rods (2 mm in diameter and spaced 15 mm) was constructed and attached to the base of Leksell stereotactic frame. The deviations between stereotactic coordinates based on magnetic resonance imaging determined in treatment planning system and real geometrical position given by the construction of an array of perspex rods within the phantom were evaluated in a series of axial and coronal images. Average deviations in all performed studies were less or equal than 0.6 mm. Phantom measurements proved minimal distortion effects for all investigated modalities and therefore no special corrections of geometrical inaccuracy were father applied. Detailed information about performed measurements can be found in [5]. Since it was decided to use polymer-gel dosimeter which is after irradiation much more sensitive to changes in transverse NMR relaxation rate R2 than longitudinal relaxation rate Ri, it was necessary to evaluate only the relative and absolute precision of transverse relaxation rate R2 for our NMR scanner. For the assessment of this precision specially made gel-dysprosium phantoms measured previously on calibrated experimental relaxometry unit in the National Institute of Health (NIH), Neuroimaging branch (Bethesda, Maryland, USA) were used. A very good relative precision of R2 2 measurement was proved by linear dependence (R = 0.9987) of R2 relaxation rates on the dysprosium concentration in the gel phantoms. There were compared absolute values of R2 relaxation rates measured on our NMR unit with those measured on experimental unit at NIH. A 5 % difference in absolute values of R2 relaxation rates was found. However, for gel dosimetry the most important parameter is the relative precision, since we want to measure a relative dose distribution. Measurements of geometrical precision and R2 relaxation rates precision proved that our NMR scanner can be used for evaluation of gel dosimeters. 21" RHD Jasná pod Chopkom 91 Only preliminary measurements of polymer-gel dosimeters irradiated by different homogenous doses (0-28 Gy) in 60Co gamma cell unit were carried out so far. Results are presented in Figure 2. 5,0 4,5 4,0 3,5 Í7 3,0 A 2,5 S 2,0 1,5 1,0 0,5 0,0 0 5 10 15 20 25 30 Absorbed dose in gel dosimeter [Gy] Figure 2 Dependence of transverse NMR relaxation rate R2 on the homogenous dose absorbed in gel dosimeter. REFERENCES 1. Gore J.C., Kang Y.S., Schulz R.J. Measurement of radiation dose distributions by nuclear magnetic resonance (NMR) imaging. Phys. Med. Biol. 29:1189 - 1197, 1984 2. Appleby A., Christman E.A., Leghrouz A. Imaging of spatial radiation dose distribution in agarose gels using magnetic resonance. Med. Phys. 14:382-384, 1987 3. Olsson L.E., Petersson S., Ahlgren L., Mattsson S. Ferrous sulphate gels for determination of absorbed dose distributions using MRI technique: basic studies. Phys. Med. Biol. 34:43-52,1989 4. Maryanski M.J., Gore J.C., Kennan R.P., Schulz R.J. NMR relaxation enhacement in gels polymerized and cross-linked by ionising radiation: a new approach to 3D dosimetry by MRI. Magnetic Resonance Imaging 11:253-258,1993 5. Novotný Jr., J.; Novotný, J.; Vymazal, J.; Liščák, R.; Vladyka, V. Assessment of the accuracy of stereotactic target localisation using magnetic resonance imaging - a phantom study. Journal of Radiosurgery 1:99-l 11,1998 92 SK98K0349 2ŕRHD Jasná pod Chopkom RADIODIAGNOSTIC MEASUREMENTS RADIATION LOAD OF CHILDREN BY CHEST D. Nikodémova2) ,M.Ranogajec2) , M. Vladár1}, M.Horváthová 3) !) Institute of Preventive and Clinical Medicine, Bratislava 2* Ruder Boškoviě Institute, Zagreb, Croatia 3> Faculty of Nursing and Social Care, Trnava, Slovakia INTRODUCTION European Communities have adopted in the 1997 the Directive 97/43 "On health protection of individuals against the danger of ionizing radiation in relation to medical exposure"[l]. Article 9 of this Directive gives special attention to the control measures, patient doses and quality assurance programmes for radiodiagnostic procedures of children. Examinations of children (aged 0 to 15 years) need special considerations in view of the longer opportunities for expression of induced cancers and of higher risk factors for certain types of Ca. The proposed reference entrance surface dose for chest radiodiagnostic examination of a 5 year old child is 100 uGy (V = 60-80 kV, HVL = 2,7 mmAl, FSD = 1,5 m). With the aim of optimalization of the radiographic examinations in paediatric radiology we surveyed the entrance surface doses measured by chest examinations, as the most frequent procedure for children in Slovakia. In our paper we present the results of measurements and calculations of doses obtained by paediatric patients with ages ranging from newborn to 15 years, who undergone chest radiodiagnostic examinations during the year 1996. The obtained results are compared with dose assessment procedures based on total absorbed dose calculations using the published conversion factors between energy imparted to the patient and entrance surface doses integrated over the beam area. MATERIALS AND METHODS Our data were collected for 6 paediatric radiodiagnostic departments in the county of Trnava, where 590 996 people are living including 188 974 children aged to 15 years. 39 096 children undergoing radiographic examinations in 1996 were split into five categories: 0-1 years, 1-4 years, 5-9 years, 10-14 years, 15 years. In the table 1. are presented the distributions of various examinations in the chosen age categories for patients undergoing radiographic examinations on the investigated workplaces. For entrance surface dose measurements individually calibrated Harshaw TLD- 700 detectors were used. The detectors were packed in plastic holders and calibrated at the Slovak Metrological Institute using the standard X-ray equipment and the exposure by 50 kV, 2 mA and 3 mmAl filtration. The TLDs were annealed for 60 min at 400 °C followed by 120 min at 100°C in an automatic microprocessor controlled PTW TLDO oven. All readings were made with computer controlled HARSHAW 3500 reader. The readout temperature was 300°C. For the phantom measurements an anthrophomorphic heterogeneous tissue-equivalent phantom was used. It corresponds to the age of 1-3 years, the height was 65 cm, the weight 10 kg. In this phantom real skeleton of baby is used and the soft tissue is imitated by curing mixture. The phantom body consists of parallel layers 2,5 cm thick with 5mm holes for inserting detectors. 95 detectors are placed inside the phantom at different organs and body parts [2]. The dimensions of the 2 ŕ RHD Jasná pod Chopkom 93 phantoms for other age categories were chosen using the published data of Persliden [3]. Table 1. Number of examinations of children in Trnava county in 1996 Organ / Age (yr) 0 1-4 5-9 10-14 15 Total Lung 1094 3397 2935 3071 548 11045 Paranasal sin. 3 1033 2731 2678 492 6937 Chest 31 146 303 855 240 1575 Extremities 190 1472 3254 6620 1244 12780 Abdomen 69 135 139 154 33 530 Skull 42 631 813 656 153 2295 Coxae 2228 999 237 151 10 3625 Others 59 53 75 86 36 309 Total 3716 7866 10487 14271 2756 39096 RESULTS AND DISCUSSION For stochastic health risk estimation associated with irradiation during radiographic X-ray examinations, obviously the effective dose, E, is used [4]. Determination of effective doses for radiologie examinations is difficult, time consuming and due to different sizes of paediatric patients also uncertain.The correlation of the concept of energy imparted to the patient to radiation risk seems to be ( especialy in the case of paediatric radiology) just as food as of more complicated estimate of effective dose. Tissue weighting factors, WT, indispensable for effective dose calculations, have not been published until now for children and the data given for adults are not generally applicable to children. Furthermore as compared to E, the energy imparted to the patient and the mean absorbed dose , Ea, will not change with time as knowledge of radiosensitivities of organs increases and are less dependent on exact knowledge of field position and field size. For this reason in the present work we used the dose absorbed during the radiodiagnostic examinations of children, given by the following equation: Ea = (ESDxAxCa)/M where are: ESD = the measured entrance surface dose in mGy A = the irradiation beam area in m2, 2 Ca = the conversion factor (kg/m ) between the imparted energy and dose area product [3], M = the mass of patient in kg. The values of Ca were calculated taking into account the age of the patient and the total filtration (HVL in mm of Al) applicated during the examination. In the table 2 technical parameters and mean ESD values for chest examinations of children of the age group 5- 9 years at various radiodiagnostic departments. The used filtration was 3 mm Al. Mean entrance surface doses measured on the surface of children by TLD, during the chest PA radiodiagnostic examinations are given in table 3, for various age categories. 94 21" RHD Jasná pod Chopkom Table 2. Technical parameters and entrance surface doses for X-ray examinations of chest (children 5-9 years) Used Technical parameters FSD ESD Hospital equipment kV mA.s cm mGy Skalica chiralux 2 38-46 24-42 100-150 0.59 Senica chiralux 2 50-53 18 100-150 0.56 Hlohovec MP 15-chirana 55-60 4.0-6.0 150 0.40 Trnava chiralux 2 50 18 150-200 0.43 Piešťany chiralux 2 50 13-18 150 0.44 Galanta durolux 71-73 5.0-12.0 150-200 0.43 Table 3 .Entrance surface doses (ESD) and absorbed doses (Ea) as a function of age for PA chest examinations Age range No. of Mean ESD St.deviation Es (years) measur. (mGy) (mGy) (mGy) 0-1 18 0.46 0.10 0.285 1-4 16 0.46 0.05 0.263 5-9 34 0.47 0.11 0.244 10-14 18 0.52 0.12 0.239 15 12 0.54 0.14 0.229 Variations in the measured ESD related to phantom sizes representing the ages 1, 3, 7,12and 15 years, can be seen in figure 1. <- - J H- -• .' Figure I. Variation of ESD as a function of calculated phantom sizes representing children ages Another possibility for Ea estimation is given by Huda and Gkanatsios [5], taking into account the relation between Ea and energy imparted, e, to the patient (or phantom) where s is equal to product of ESD x A x Ca. Ratio of absorbed dose Ea to energy imparted s as function of the mass of patient is demonstrated infigure 2 . 2 ľ RHD Jasná pod Chopkom 95 asi _1I1f!• "••• '1.'.. i1I\•Bili mg -= = Sa z mm S - -~» 1 * g; Š? S !r X?tn i « ^aó/ 1•y m Figure 2. Absorbed dose vs. phantom mass for one Joule of phantom irradiation CONCLUSION The procedure described in our study allows to estimate the radiation load of children during radiodiagnostic examinations. One difficulty connected with the used methods is that the conversion factors are derived only for limited number of energy distributions. Radiation quality given by HVL provides only a rough estimate of the distribution of energies actually used for examination of the patient. To avoid this problem we will stream to extent this work by experimental measurements of the depth- doses curves in the phantom simulating the body of the child [6]. In this way we can derive the conversion factors between entrance surface dose and the mean absorbed dose to the organs (instead of total absorbed dose Ea). We are expecting that from the standpoint of radiation risk assessment more exact results should be obtained. REFERENCES 1. Council directive 97/43/Euratom. On health protection of individuals against the danger of ionizing radiation in relation to medical exposure. 1997 2. Gubatova D, Varchenya V, Krastinya A. Tissue-equivalent phantoms in radiological protection. Kernenrgie 1989; 32: 10-13. 3. Persliden J, Sandborg M. Conversion factors between energy imparted to the patient and air collision kerma integrated over beam area in paediatric radiology. Acta Radiologica 1993; 34: 92-98. 4. Almen A, Nilsson M. Simple methods for the estimation of dose distributions, organ doses and energy imparted in paediatric radiology. Phys. Med. Biol. 1996; 41: 1093- 1105. 5. Huda W, Gkanatsios N.A. Effective dose and energy imparted in diagnostic radiology. Med. Phys. 1997; 24: 1311-1316. 6. Nikodémova D, Horváthova M, Ranogajec M. Estimation of radiation load of children during chest radiodiagnostic examinations. Slovak Radiology. 1998; 5: 57. 96 SK98K0350 21"RHD Jasná podChopkom RADIODIAGNOSTIC MEASUREMENTS RADIATION LOAD OF CHILDREN BY CHEST D. Nikodémova1} ,M.Ranogajec 2), M. Vladár1}, M.Horváthová 3> ' Institute of Preventive and Clinical Medicine, Bratislava 2) Ruder Boškoviě Institute, Zagreb, Croatia 3) Faculty of Nursing and Social Care, Trnava, Slovakia EXTENDED SUMMARY Radiodiagnostic examinations of children present particular importance from the radiation hygiene point of view. The estimation of the radiation load of paediatric patient is not easy, because of the lack of information about organ weighting factors for various ages of patients, as well as due to the differences in applied X-ray examination parameters. In the district of Slovak Republic, in which also the working NPP Jaslovské Bohunice is included, we have tried to estimate the radiation load of children to 15 years by chest radiodiagnostic examinations. Our data of entrance surface doses were collected using measurements with TLD for 100 patients divided in 5 age categories at six radiodiagnostic departments. The calculations of the total absorbed dose were performed using the measured ESD values integrated over the X-ray beam area, the conversion factors between the imparted energy and the dose-area product and the known irradiation parameters (kV, HVL, mass, etc.). The analysis of the obtained absorbed doses (Ea) as a function of age for chest PA radiodiagnostic examinations has shown, that the investigated Slovak radiodiagnostic centres use rather lower voltage techniques and the entrance surface doses are much higher than the proposed value of EC. 21" RHD Jasná pod Chopkom SK98K0351 97 RADON EXPOSURE AND LUNG CANCER RISK Ladislav Tomášek Státní ústav radiační ochrany, Šrobárova 48, 100 00 Praha Research supported by the Internal Grant Agency of the Ministry of Health (IGA 4920-3). Introduction Epidemiological evidence of lung cancer risk from radon is based mainly on studies of men employed underground in mines where exposures are relatively high in comparison to indoor exposure. One of first such studies was established by the late Josef Sevc. In 1971, the first results of this large epidemiological study were reported [1]. During the next 20 years, Ševc published a number of results based on extended follow-up of the original cohort and established two further cohorts. The series of his papers closed in 1991 [2]. In 1990, Ševc designed and initiated a new large study among inhabitants exposed in homes, which was then the only cohort study of indoor exposure. Occupational exposure in mines The studies of miners involve three cohorts (Tab.l). The oldest cohort (S) comprises uranium miners at the Jáchymov region firstly exposed in 1948-59, miners of the second cohort (N) entered the Poibram uranium mines in 1968-74, and the third cohort (L) includes burnt-clay miners in the Rakovník district employed in 1960-80. Exposure estimates in the S study were derived from large numbers of measurements of radon. Each man's annual exposures to radon progeny were estimated combining measurement data with the men's registered employment details [3]. In the N and L study, the exposure estimates were mostly based on personal dosimetric records [4]. The statistical analyses were based on relative risk models in the general form O = cE(l+ERR(W,Z)), where O denotes the observed number of cases, E is the number expected from national mortality data, ERR is the excess relative risk function depending on exposure W and modifying variables Z, and c is an intercept term that allows the mortality rate for the 'unexposed1 cohort to differ from that in the general population [3]. The most simple model of relative risk assumes a linear dependence on total cumulative exposure W lagged by 5 years. The coefficient of proportionality - the excess relative risk per WLM (ERR/WLM) is called the relative risk coefficient. The effect of time and age related modifying factors is studied by the method of exposure windows which consists in separating the individual exposure into segments corresponding to different periods and age categories. Tab. 1. Czech cohorts status by 1995 Cohort Since Size Cases O/E Died Mean exposure S 1952 4320 788 4.76 66% 155 WLM N 1969 5630 54 1.48 11% 7 WLM L 1960 914 66 2.19 45% 28 WLM 98 2 ľ' RHD Jasná pod Chopkom By 1995, a total of 908 cases of lung cancer were observed in the three cohorts (Tab.l). In the simple uniform model of relative risk (Fig.l), the ERR/WLM estimate for all studies combined is 0.016 (95%CI: 0.011-0.021). Risk coefficients for separate studies - 0.012 (S), 0.057 (N), 0.019 (L) reflect differences in the cohorts, namely the length of follow-up and age at exposure. For instance, in the N cohort time since exposure is about 20 year shorter and age at exposure is lower of nearly 10 years. — ERR/WLM 0.016 100 200 300 400 WLM Fig. 1. Relative risk by cumulative exposure among miners With the extended follow-up, factors that modify general linear model become more discernible. The recent analyses demonstrated significant modifying effects of: (1) time-since-exposwe (TSE), which decreases the ERR/WLM to less than 10% after 25 years since exposure, (2) age-at-exposure (AE), with two-fold ERR/WLM for age before 30 and half- coefficients for age after 40, (3) exposure rate, which reduces the ERR/WLM to 30% for exposure rates higher than 8WL. The parameters of the complex model translated into age-at-exposure and time- since-exposure specific ERR/WLM coefficients illustrate temporal and age related changes in the magnitude of risk coefficients (Tab.2). Tab.2. TSE&AE specific risk coefficients at low exposure rates Studies of miners are at present the principal source of information on the effects of exposure to radon. Radiation protection limits in uranium industry, derived already in 2 ľ' RHD Jasná pod Chopkom 99 1972 from the S study, were adopted by some other countries. The S study contributed substantially to the recommendations of the International Commission on Radiological Protection (ICRP-65: Protection against radon-222 at home and at work, 1993). The recent quantitative results of the studies significantly improved the compensation scheme presently used for ex-miners. Indoor exposure The indoor study is designed as a prospective cohort follow-up covering period since 1960. The study area - Middle Bohemian Pluton - is mostly granitoid with considerable breaks. The study population include inhabitants of the area (80 villages) who had resided there for at least 3 years, alive by the end 1960 or born later. A total of 12 010 satisfied these criteria. The population is characteristic by low migration. By 1995, almost quarter of the cohort have died, 168 from lung cancer (Tab.3). As some causes of death have been still missing, the present figures suggest slightly increased mortality from lung cancer in comparison to generally low numbers corresponding to cancers other than lung. Tab.3: Cause and Period Specific Mortality Period Lung NLung Violent Other Unk All Deaths Ca Ca n PY 0 O/E 0 O/E 0 O/E 0 O/E 0 0 O/E 1961-70 69 226 30 0.82 121 0.85 34 0.59 569 1.06 14 768 0.99 1971-80 77 367 60 1.28 154 0.89 49 0.71 679 0.93 11 953 0.94 1981-90 81 101 62 1.28 154 0.87 510.76 796 1.08 14 1077 1.05 1991-95 41 048 16 0.67 66 0.75 22 0.62 289 0.96 9 402 0.90 Total 268 742 168 1.08 495 0.85 156 0.68 2333 1.01 48 3200 0.98 The exposure assessment was based on one year measurements in houses (2500) in the study area (Kodak LR115). Exposure estimates in residences outside the area were derived from a large scale mapping of radon in the country [5]. The individual exposures in terms of kBq m'3a were estimated combining the radon equivalent equilibrium concentration (Bq tri3) and duration of residence (a). For houses in the study area that were not available for measuring, the community means were used instead of missing values. Concentrations corresponding to residences outside the study area (21%) were estimated by (1) larger community means for inhabitants in the neighbouring four districts and by (2) district means for the residences in other districts, where concentrations were usually much lower (Tab.4). These numbers reflect relatively low migration in the study population. The distribution of the cumulative exposure in the cohort was considerably asymmetrical, 79% of individual exposures were below 20 kBq m'3a with mean 14 kBq m'3a and maximum 647 kBq m'3a. In evaluating exposure - response relationship, last 5 years were ignored, as well as the exposures before more than 25 years. The relative risk (O/E) was related to exposure in time window 5-25 years previously (Fig.2). The observed number of cases is not definitely large enough to give reasonably precise conclusions on the risk 100 21" RHD Jasná pod Chopkom coefficient, nevertheless the present estimate of the ERR/kBq m*3a - 0.049 ( 95% CI: 0.018 - 0.079 ) is statistically significant. Further efforts are directed to improve exposure estimates by measuring missing houses and by extending follow-up. Tab.4: Estimation of Radon Equivalent Equilibrium Concentrations PYofExp % Bq/m3 Direct Measurements 377 084 65 324 Community Means 84 277 14 312 Neighbouring Districts Communities 59 691 10 181 District Means 61699 11 90 Total 582 751 100 282 0/E — 0.049 , , . . I . . . . I , . , . I . . , . I . . . . T 95% CI 10 15 20 kBq/m3 a Fig2: Relative risk by cumulated exposure in the indoor study Note One Working Level (WL) equals any combination of radon progeny in one litre of air which results in the ultimate emission of 1.3 105 megaelectronvolts (MeV) of energy from alpha particles. The WLM is time-integrated exposure measure, i.e. the product of time in working months (170 hours) and working levels (1WLM=3.54 mJh m'). References [1] Ševc J, Plaček V, Jeřábek J. Lung cancer risk in relation to long-term radiation exposure in uranium mines. Proc. 4th Conf. on Radiation Hygiene, Jasná pod Chopkom, Part E: 315-326,1971. 21st RHD Jasná pod Chopkom 101 [2] Ševc J, Tomášek L, Plaček V. Riziko zhoubného novotvaru plic pei inhalaci dce0iných produktu radonu (Risk of malignant lung tumours as a result of inhalation of radon daughter products). Ěs.hygiena 36:3-13,1991. [3] Tomášek L, Darby SC, Fearn T, Swerdlow AJ, Placek V, Kunz E. Pattern of lung cancer mortality among uranium miners in West Bohemia with varying rates of exposure to radon and its progeny. Radiat. Res. 137:251-261,1994. [4] Plaček V, Tomášek L, Heribanová A, Kunz E. Zhoubný novotvar plic a expozice radonu v souěasných podmínkách důlního provozu (Lung cancer and exposure to radon under present mining conditions). Prac.lékaaství 49: 14-20,1997. [5] Hůlka J, Fojtíková I, Borecký Z, Tomášek L, Burian I, Holeček J, Thomas J. Indoor radon risk mapping in the Czech Republic. Proc. Europ. Conf. on Protection against Radon at Home and at Work, Prague, June 1997. 102 SK98K0352 21st RHD Jasná pod Chopkom RESULTS OF CYTOGENETIC EXAMINATIONS OF MINERS EXPOSED TO RADON IN ORE MINES Beňo, M., Vladár, M., Nikodémova, D., Vtčanová, M. Institute of Preventive and Clinical Medicine, Limbová 14, SK-83301 Bratislava 1. Introduction Radon causes lung cancer in the miner population occupationally exposed underground to mine air of uranium mines [1-3]. There is increasing evidence supporting the opinion that radon may cause cancer also in ore mines [4-7] despite that in ore mines the radon gas is released in concentrations down to several orders of magnitude lower than in uranium mines. At these low concentrations of radon other pollutants may play increasing role in carcinogenesis and clastogenesis, respectively. There were reports about higher incidence of lung cancer also from ore mines of Slovakia and the radon exposure underground was suggested to be the main cause of this higher incidence [8-10]. The chromosomal aberrations (CA) are considered to be indicators of irradiation and evidence rises that they belong to important factors of somatic mutagenesis leading to cancer [11]. Specific localizations of CA are found in many cancer tissues including lung cancer [12,13]. A positive dependence of numbers of CA in peripheral blood lymphocytes from lifetime underground exposure in uranium ore miners has been reported [14]. Similarly, a dependence of numbers of CA in peripheral blood lymphocytes from lifetime exposure to indoor radon at very low levels has been observed in inhabitants of dwellings [15]. The above information has lead us to perform a study in which the radon air concentrations and clastogenic effects at three ore mines located in central east Slovakia, the gold mine of Hodrusa-Hamre, talcum mine of Hnusta, and, iron ore mine in Nizna S lana are compared with the above parameters observed in a control group of healthy men which never experienced underground work. A random sample of radon concentration mesurements in houses was used for control. 2. Materials and Methods Samples of venous blood from groups of 21, 32, 38 and 39 probands from Hodrusa-Hamre, Hnusta, Nizna Slana and from healthy control persons, respectively, were taken after an informed consent. The average age of the groups did not differ significantly. Every proband was interviewed and a questionnaire was filled up which helped to exclude probands with diseases requiring frequent x-ray examinations, recent viral diseases, exposure to chemicals or therapeutic drags, or, excluding those having previously worked in uranium mines. Information about smoking and drinking habits was gained. Cultures from one-hour buffy coat were started using RPMI1640 (USOL Praha) medium supplemented with phytohaemagglutinin (MUREX, 0.025 ml/ml), antibiotica (Penicillin BIOTIKA, 100 i.U./ml, Streptomycin GALENIKA, 100 microg/ml), and 10% fetal calf serum at 37° C. The cells were harvested after 48 hours of cultivation, for last two hours 4 ug/ml colchicine (FLUKA) was added. The cells were then treated with 0.075 M KC1, fixed in methanol : glacial acetic acid (3:1) and dropped on wet slides. All slides were coded and stained in 3% Giemsa solution. Chromosomal aberrations were counted in 200 well stretched 2 ľ1RHD Jasná pod Chopkom 103 mitoses from every proband and recorded separately as chromatid and chromosomal fragments, breaks, terminal deletions, chromatid and chromosomal gaps, dicentrics, and chromatid exchanges. Personal doses were measured by a pair of solid state nuclear track detectors of type CR-39 in a passive two-chamber system monthly, during 1995 year. Electrochemical etching combined with chemical pre-etching was used for evaluation the track detectors (16). The same dosemeters were used to assess the radon concentration in homes. For biometrical evaluation the two-times-two table of chi-quadrat and at low aberration frequencies the Fisher's exact test of the program package EPIINFO, and, for confidence interval estimates the program BINOM [17] was used. 3. Results The proportions of mitoses with aberrations of the chromosomal type in samples from all followed groups are in Table 1. It can be seen that the variation between means of these proportions generally does not exceed a factor of two. If comparing the total (non-smoker plus smoker) samples a statistically significant (p<0,05) difference could be seen between the counts of control and Hodrusa samples on one side and Nizna Slana samples on the other. The responsibility for this difference lied upon the samples of Nizna Slana smoker-miners as could be seen from the statistically significant difference at comparing smoker subgroups. No significant differences were found among other sample groups. Table 2 shows the averages of sums of all cells containing CA including gaps. The variation between groups does not exceed the factor of two. A statistically significant difference between the miners and control group samples was found. Also here, mainly the miner smokers subgroup contributed significantly to this difference. In comparing with the counts of cells containing C A including gaps in samples from smoking controls significantly higher counts in smoking miners (p<0,05) were found. A prominence of Nizna Slana smokers is evident. In comparing the counts in control and miner non-smokers the difference was not statistically significant. From the experiments of Purrott et al. [18] who, after in vitro alpha irradiation of cells, found an overdispersion of CA in comparing with the Poisson distribution it follows, that the probability of the occurrence of cells with more than one CA is after alpha irradiation higher than after gamma irradiation. Therefore special attention was paid to cells containing more than one aberration as these might be markers of effects of alpha radiation. Generally, no significant difference in counts of cells containing more than one CA either of chromosome type, or, of chromatid type in comparing with the control samples was found. Table 3 shows the average concentrations of radon in air and resulting annual average radiation body burden estimates from radon at the followed mines and in two types of houses, appropriate for the control group, the majority of which lodged at rack construction houses. As expected, the annual effective dose estimates were higher for miners than for controls. The dose estimates between controls and miners varied by a factor of 5 to 10, depending from the housing quality, and respective mine. The highest value of effective dose was estimated for the mine Hodrusa-Hamre, but owing to high vartiation no statistically significant differences between annual effective dose estimates within the three mines were found. 4. Discussion 104 21" RHD Jasná pod Chopkom According to contemporary criteria only the counts of cells with CA of the chromosomal type reflecting the actions of clastogens during the Go to Gi phase of the lymphocyte cell cycle are used for the in vivo clastogenic effect assessment. This, because the overwhelming majority of circulating lymphocytes are non- cycling cells accumulating and steadily repairing the damage caused by exposure to clastogens. The results of this study show a statistically significant clastogenic effect in Nizna Slana smoker miners having higher aberration counts in comparing with non-smoker controls and with their non-smoker co-workers. This result and also the outcome of evaluating the all-aberration counts strongly suggests that interaction of the working environment in this mine with the smoking habit might contribute to this finding. The range of the differences between the effective dose- estimates of control and miner groups within a factor of 5 to 10 as compared with the differences in the CA counts which vary within a factor of about two shows that the relation of the clastogenic effects to the underground exposure may not be ruled by a simple proportionality. The role of smoking as an agent aggravating the clastogenic effect of the underground environment is demonstrated by the absence of significant differences in clastogenic effects among non-smoker sample groups. Agostini et al. [19] who, not accounting for exposure to radon, sought clastogenic responses to the underground environment of coal miners found a significantly higher count of cells with at least one chromosome alteration. They also found higher counts of cells with CA if they included gaps and concluded that an unknown substance in serum of the miners may have caused the higher occurrence of gaps as they found significantly elevated counts of gaps, fragments and overall alterations in preparations made from control lymphocytes cultivated in miner plasma. This may be true also for the samples of the miners, especially for the smokers subgroup in this study. The information [19] came late so that in this study no attempt has been made to assess such substance in plasma. In connection with significantly higher counts of gaps attention should be drawn to recent in vitro experiments with endonucleases elucidating the variability of gap-counts and indicating that "gaps represent localized difficulties in chromatin condensation due to unrepaired DNA double strand breaks" [20]. Accounting for an uninterrupted repair of chromosomal damage proceeding in the cells up to the fixation, hypothesised substances in serum of smokers may contribute to prevent the repair of gaps leading to significant increase of gap counts as in preparation of standard cytogenetic samples lymphocytes in a small volume of the probands plasma are added to the cultivation medium. Smoking invariably acts as an in vivo clastogenic factor [21], moreover, Vijayalaxmi et al. [22] found higher counts of gaps in samples of lymphocytes from smokers. The nature and in vivo action of these substances remain to be cleared. More clastogens, as diesel exhaust fumes, mineral dust containing clastogenic metal compounds, mycotoxins [23] etc. present in the environment of mines might contribute to the appearance of such substances in blood plasma, however, the exposure to these factors were not measured in this study. In order to assess the clastogenic risk of underground radon more precisely in future it is necessary to correlate the endpoints of clastogenesis to the estimates of radiation exposures measured by personal dosimeters similarly as at measuring exposures to indoor radon. Nevertheless, this study indicates that the clastogenic risk from underground professional exposure to such low levels of radon as 2ľ RHD Jasná pod Chopkom 105 encountered in ore mines, may be effectively confounded by that from smoking. To elucidate this problem higher counts of probands in non-smoker and smoker in both control and miner groups should be examined. 5. Conclusions Significant differences in counts of aberrations of the chromosomal type in lymphocytes of smoker-miners of Nizna Slana as compared with counts of such aberrations in lymphocytes of a control group of similar age were found. A dependence of chromosomal aberration counts from the underground exposure to radon by multiple regression procedures could not be ascertained. The results indicated that confounding of such dependence by smoking might have taken place. To elucidate this problem higher counts of probands in non-smoker and smoker in both control and miner groups should be examined and the personal dosimetry of radon should be used for assessment of individual radiation body burden. Significantly higher counts of gaps in the miner samples were found and it could not be excluded that, hypothetically, they might arise as a result of in vitro exposure to some unknown substance(s) present in smoker miners plasma introduced into samples together with lymphocytes at the beginning of their cultivation and inhibiting the DNA repair processes. 6. References [1] J. Sevc, E. Kunz, V. Placek, Health Phys. 54 (1988) 27-46. [2] J. Sevc, L. Tomasek, E. Kunz, et al, Health Phys. 64 (1993) 355-369. [3] L. Tomasek, A.J. Swerdlow, S.C. Darby, et al. Environ. Med., 51 (1994) 308-315. [4] K.G. StClair Renard, Ambio 2 (1974) 67-69. [5] V. Pekarek, C. Lazar, J. Rýchlik, Pracov. Lek. (in Czech) 34 (1982) 270-272. [6] S.C. Darby, E.P. Radford, E. Whitley, Environ. Health Perspect. 103 Suppl 2 (1995) 45-47. [7] E.P. Radford, K.G. StCIaír Renard, New Engl. J. Med. 310 (1984) 1485-1494. [8] J. Buchancova, A. Zigova,etal., Folia MedicaMartiniana, 15(1988)253-259. [9] J. Buchancova, A. Zigova,, Stud, pneumol. et phtis. Cechoslov., 48 (1988) 512-525. [10] J. Icso, M. Szollosova, Pracov. Lek. (in Slovak) 36 (1984) 294-298. [11] UNSCEAR , Radiation carcinogenesis in man. In : Sources, effects and risks of ionizing radiation. United Nations Scientific Committee on the Effects of Atomic Radiation Report to the general Assembly, United Nations, New York, (1988) 405-543. [12] Y.E. Miller, W.A. Franklin, Hematol. Oncol. Clin. North America, 11 (1997) 215-234. [13] J.S. Wiest, W.A. Franklin, H. Drabkin, et al., J. Cell. Biochem. Suppl, 28/29 (1997) 64-73. [14] W.F. Brandom, A.D. Bloom, P.G. Archer, V.E. Archer, R.W. Bistline, G. Saccomanno, Somatic cell genetics of uranium miners and plutonium workers. A biological dose-response indicator, in: Proc. Symp. on Late Biological Effects of Ionizing Radiation, Vol.1., IAEA Vienna, 1978, pp.5O7- 517. [15] M. Bauchinger., E. Schmid, et al., Mutation Res. 310 (1994)135-142. [16] M. Vicanova, M. Durcik, D., p. 195-198 in: Proc. IRPA Regional Symp. on Radiation Protection, Prague, Czech Republic 1997. [17] Mikulecky, M., Komorník, J., Ondrejka, P. BINOM - Automation of mathematic-statistical estimates and tests on the basis of binomial distribution. Comtel, ed., Bratislava, 1998. [18] BJ. Purrott, A.A. Edwards, D.C. Lloyd, et al., Int. J. Radiat. Biol. 38 (1980) 277-284. [19] J.M.S. Agostini, P.A. Otto, A. Wajntal, Brazilian J. Genet. 19 (1996) 641-646. [20] A.N. Harvey, N.D. Costa, J.R.K. Savage, et al., Somat. Cell Molec. Genet. 23 (1997) 211-219. [21] G. Obe, H.J. Vogt, S. Madle, A., et al., Mutat. Res. 1982; 92: 309-319. [22] Vijayalaxmi, H.J. Evans, Mutation Res. 1982; 92:321-332. [23] R. Sram, L. Dobias, P. Rossner, et al., Environ. Health Perspect. 101 Suppl.3 (1993) 155-158. [24]International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, Safety Series 115, IAEA, Vienna 1996 106 2ľ RHD Jasná pod Chopkom Table 1. Percentages of mitoses containing aberrations of chromosomal type in control persons and miners of three Slovak mines Total Non-smokers Smokers n mean 95%CL n mean 95%CL n mean 95%CL Control 7800 1.10* (0.9-1.4) 4200 1.02° (0.7-1.4) 3600 1.19* (0.9-1.6) Hodruša 4200 1.04a (0.8-1.4) 1400 0.92 (0.5-1.6) 2800 l.lla (0.8-1.6) Hnúšťa 6400 1.41 (1.1-1.7) 3200 1.25 (0.9-1.7) 3200 1.56 (1.2-2.1) a -abc N.Slaná 7600 1.63* (1.4-1.9) 2400 1.04" (0.7-1.5) 5200 190 (1.6-2.3) Legend: Total = non-smokers + smokers; n = number of evaluated mitoses; CL = confidence limits *" within columns shows statistically significant (p<0.05) differences between samples b c across columns shows statistically significant (p<0.05) differences between samples Table 2. Percentages of mitoses containing chromosomal aberrations including gaps in control persons and miners of three Slovak mines Total Non-smokers Smokers n mean 95%CL n mean 95%CL n mean 95%C1 Control 7800 5.43>a (4.9-6.0) 4200 5.78* (5.1-6.5) 3600 5.02"a (4.3-5.1 Hodruša 4200 7.07*b (6.3-7.9) 1400 6.43 (5.2-7.8) 2800 7.39*a (6.5-8.' Hnúšťa 6400 5.58 te (5.0-6.2) 3200 5.19s (4.4-6.0) 3200 6.41* (5.6-7.: N.Slaná 7600 8.18* (7.6-8.8) 2400 6.88 (5.9-8.0) 5200 8.78<&s (8.0-9.( Legend: Total = non-smokers + smokers; n = number of evaluated mitoses; CL = confidence limits "ai>c within columns shows statistically significant (p<0.05) differences between samples; * s across columns shows statistically significant (p<0.05) differences between samples; Table 3. Results of radon measurements and effective dose estimates in Slovak houses and mines Houses (Bratislava)" Mine0) Parameter Unit Family Rack Total Hodruša Hnúšťa N.Slaná houses construction No. of samples [1] 130 357 487 68 129 513 Arith. mean [Bq.rn3] 68 13 28 999 789 698 Minimum [Bq.m3] 3 3 3 85 77 53 Maximum [Bq.m-3] 905 109 905 3812 4639 4688 Geom. mean [Bq.m"3] 40 10 14 721 550 478 STDEV(Geom) [Bq.m-3] 1,7 1,3 1,8 2,4 2,5 2,4 [mSvl 2,9 0,6 1,2 6,2 4,9 4,4 a) EEC b) CuaRadon concentration in air c) E = arithmetic mean x Conversion factor. Conversion factor [mSv/Bq.m"3] = 3.90 E-2 for houses and 6.23 E-3 for mines [24] 21st RHD Jasná podChopkom SK98K0353 107 Analysis of the similarity factors of the villages in the areas of the nuclear power plants from the premature death-rate performed by fuzzy logic method. M. Letkovičová, R. Řehák, J.Korec, B. Mihály. VI. Príkazský, Environment corp., CBE, Nitra, Slovak Republic Nuclear Power Plant Research Institute Trnava corp. Trnava, Slovak Republic Slovak Electric Power Plants corp., Bratislava, Slovak Republic National Institute of Hygiene and Epidemiology, CEM, Prague, Czech Republic Our paper examines the surrounding areas of nuclear power plants from the proportion of premature death-rate, which is one of the complex indicators of the health situation of the population. Specifically, we focused our attention on nuclear power plant in Jaslovske Bohunice (SE-EBO), which has been in operation for the last 30 years and nuclear power plant Mochovce (SE-EMO), which was still under construction when data was collected. WHO considers every death of the indidvidual before 65 years of age a premature death case, except death cases of children younger than 1 year. Because of the diversity of the population, this indicator is a standard for the population of Slovak Republic as well as for the european population. The objective of the work is to proove, that even a long term production of energy in nuclear power plant does not evoke health problems for the population living in the surrounding areas, which could be recorded through analysis of premature death cases. This comes from a preassumption, that if there is a negative influence of the nuclear power plant on the population living in the area of it and its health, there should be premature death cases in the whole area or in certain geographical directions from the nuclear power plant, for example in the direction of main wind streams or in the villages on the banks of the local water streams, that ensure the float of the used water from the power plants. A particular situation has been evaluated in the individual villages, being in zones of perimeter 5, 10 and 30 kilometers from the nuclear power plant. We selected the region of the city Galanta as a control region, because it fullfills all the conditions : it is far enough from both of the nuclear power plants, there is a similar geographical situation and similar employement rate of the population, and it is not influenced by the main wind streams coming neither from SE-EBO nor SE-EMO. The parameter is at the same time counted for the whole Slovak Republic, in order to achieve the average value for the whole country. The data is aggregated into 4 annual totals of the real inputs for the individual regions. Further computations are accomplished from 4 annual totals for the years 1993 - 1996. The premature death-rate (PSPU) is directly standardized for the european population according to the following formula: PSPU = 100 {p(105.Wj.(kj/mi).ii)] / [£{105. Wj. (k|/mj)]} PSPU directly standardized premature death-rate i age group W; european standard k; number of death cases in the village within an age group mj age group ij age coefficient The homogenity of the area from the perspective of premature death was tested by the fuzzy logic method. A term "small difference" was defined to differentiate the proportion of the premature death cases between individual villages. 108 2ľRHD Jasná pod Chopkom The function of the term "small difference" within the observed feature is: Transitional outcome of the fuzzy logic relation of the similarities enables us to perform the division into equivalence classes (alfa cuts are taken as relations of equivalency). The outcome of the division into equivalency classes through alfa cuts are classes of similar villages, where the value of alfa = 0.8 is considered high enough to claim similarity. Outcomes: We evaluated 4 regions from the Directly standarized premature death-rate (PSPU): 1. The area of SE-EBO, 5 km from the centre of the nuclear power plant (EBO - 5), pic.1 2. The area of SE-EBO, 10 km from the centre of the nuclear power plant (EBO -10), pic.1 3. The area of SE-EMO, 10 km from the centre of the nuclear power plant (EMO -10), pic.2 4. The area Galanta city (the capital of the region) up to 10 km from the centre of the town (Galanta-10) 5. Slovak Republic (SR) Region Number Populati Numb, Brutto Percentage PSPU Number of villages in the zone of on 1996 of death of Slovak average (PSPU %) villages death ratio cases aver.-min. - max <8.4 8.4-16.5 >16.5 EBO-5 7 6809 71 10.4 9.5 2.5 13.6 3 4 0 EBO-10 27 36116 403 11.2 11.7 2.5 20.4 4 20 3 EMO-10 24 25639 299 11.7 12.9 0 22.4 2 18 4 Galanta-10 24 74783 810 10.8 13.1 6 20.3 2 18 SR 2908 5378932 :;5íi236 v .9i53 , 1215; i0 Í62 . •- •;54& . •., im ; í 525-. ý. We counted the directly standardized premature death-rate of the population for each village in Slovakia for the period of 1993 - 1996. We compared the area of EBO with perimeter of 30 kilometers with the situation in all Slovak Republic. The limits for the average in Slovakia are 8.4 -16.5%. 96% of the population from the 30 kilometer zone is in the average area or below it. Only 3.5% of the population is in the above average zone. The whole population from the area of 5 kilometers from EBO is in the average or below average zone; no village was in the above average zone. The exact delineation of the existence or non-existence of accumulation of the villages with higher values was accounted through accumalation based on the fuzzy logic method. The following villages (pic.1) are in the 5 kilometer zone in the area of SE-EBO: Vefké Kostolany (PSPU=13,49%), Pečeňady (PSPU=13,60%), Rado_ovce (PSPU=2,50%), Jaslovské Bohunice (PSPU=9,41%), Ratkovce (PSPU=7,77%), Dubovany (PSPU=12,98%), Nižná (PSPU=6,74%). All the villages are similar for alfa=0,96, which is considered to be a very high level of similarity. Generally speaking the objects are considered similar if the value of alfa reaches 0.8. We can proclaim, that all the villages within the observed zone are similar from the premature death perspective. The following villages (pic.2) are in the observed zone of the area of SE-EMO: Nemčiňany (PSPU=22,35%), Malé Kozmálovce (PSPU=13,51%), Čifáre (PSPU=9,26%), Nevidzany (PSPU=9,59%), Nový Tekov (PSPU=16,31%), Červený Hrádok (PSPU=13,01%), Malé Vozokany (PSPU=13,6%), Veľký Ďur(PSPU=12,69%) and Tajná (PSPU=0,0%). 2 ľ RHD Jasná pod Chopkom 109 All the villages are similar for alfa = 0,94 and so we can proclaim as in the example above, that all the villages within the observed zone are similar from the premature death perspective. Conclusion : Using the fuzzy logic method when searching for similar objects and evaluating the influence of the nuclear power plant on its surrounding area seems more natural than classical accumulation method, which separates objects into groups. When using the classical accumulation method, the objects in particular accumulation group are more similar than 2 objects in different accumulation groups. The similarity is influenced by the chosen metric. When using the fuzzy logic method the similarity is defined more naturally. Within the observed regions of the nuclear power plants, the percentage of directly standardized premature death cases is almost identical with the average for Slovak Republic. The most closely observed region of SE-EMO up to 5 kilometers zone even shows the lowest percentage. Also we did not record any areas that would have unfavourable values from the wind streams perspective neither than from the local water streams receipents of SE-EBO Manivier and Dudváh. The region of SE-EMO is also within the Slovak Republic average, unfavourable coherent areas of premature death case are non existent. Galanta city region comes out of the comparision with the relatively worse results. So we claim, that 20 years of operation of SE-EBO does not influence the health situation in its surrounding area that would be shown in premature death cases in any of the villages close by. Bibliography: 1.Beáta Stehlíková : On the similarity in demography.XXXV. Wandertagung fur Geschichte der Statistik, Ungarische Statistische Gesellschaft, Szeksárd, 7-8. mai 1998,19-20. 2. M. Letkovičová, R. Řehák and collective: Monitorovanie stavu životného prostredia a zdravotného stavu obyvateľov okolia jadrovej elektrárne Jaslovské Bohunice, výskumná správa č. 348/97,1997. 3. R. Rehák, M. Letkovičová and collective: Monitorovanie stavu životného prostredia a zdravotného stavu obyvateľov okolia jadrovej elektrárne Mochovce, výskumná správa č. 347/97, 1997. 110 2ľ' RHD Jasná pod Chopkom Directiy standardized premature death-ratE area of the NPP Jaslovské Bohunice computationfor tti e years 1993 -1996 Area of the NPP Jaslovské Bohunice wind direction (%) Percentage of the premature death . <8,4 15.0-.. | 8,4 -16,5 (average of SR) T 3 >16,5 w sw. ' ••- ... ---•''' , • SE Pic.1 21" RHD Jasná pod Chopkom 111 Directiy standardized premature death-rate area oftfie NPP Mochovce computation for the years 1993 -1996 Area of the NPP Mochovce Percentage of the premature death wind direction {%) <8,4 N NW ' vsľ^ľ"" 15,0.. _/ \ 10.0. y. \ W V ^\ J--W^~ sw "• PiC.2 112 2ŕ RHD Jasná pod Chopkom "'"^nsÄÄ'ssÄľss; a =0,96 Vafcé Kostol, Radošovce, Jastovské Bohunta, RaKovce, D^ny, Niíná a =0,97 Velké Ktata^ny, PečeAady, OUtwaw Oslovské Bohunice ' * Ratkovce, Nižná Radošovce \ Pic.3 Chartof the similartfyfactors ofihe villages from the premature death-rate area of the NPP Mochovce (SE-EMO) a =0,94 Mate K«njJtovce,ČÍ(äre, Nevidzany, Nový Tekov, Červený Hrádok Malé Vozokany, Velký Ďur, Nemčiňar^ Ta»á Pic.4 21st RHD Jasná pod Chopkom SK98K0354 113 TLD AUDIT IN RADIOTHERAPY IN THE CZECH REPUBLIC Daniela Kroutilíková 0, Helena Žáčkova #, Libor Judas + * National Radiation Protection Institute, Šrobárova 48,100 00 Prague 10, Czech Republic + Dept. of Oncology, 1.Medical Faculty, Charles University, Kateřinská 32, 120 00 Prague 2, Czech Republic ABSTRACT Purpose: National Radiation Protection Institute in Prague organizes the TLD audit. The aim of the TLD postal audit is to provide a basic control of the clinical dosimetry in the Czech Republic for purposes of state supervision in radiotherapy, to investigate and to reduce uncertainties involved in the measurements of absorbed dose and to improve consistency in dose determination in the regional radiotherapy centers. Materials and methods: The TLD audit covers absorbed dose measurements under reference conditions for ^Co and 137Cs beams, high-energy X-ray and electron beams of linear accelerators and betatrons. The TL-dosemeters are sent regularly to all radiotherapy centers. Absorbed dose measured by TLD is compared to absorbed dose stated by radiotherapy center. Encapsulated LiF:Mg,Ti powder is used for the measurement. Deviation of ±3% between stated and TLD measured dose is considered acceptable for photons and ±5% for electron beams. Results: First TLD audit was started in 1997. A total of 135 beams was checked. There were found seven major deviations (more than ±6%), which were very carefully investigated. Medical physicists from these departments reported a set-up mistake. However, at most of those hospitals with major deviations, an in situ audit in details was made soon after TLD audit. There were found discrepancies of clinical dosimetry but also bad technical state of some of the irradiation units. In 1998, second course TLD audit was started. No major deviation was found. Conclusions: Regular TLD audit seems to be a good way to eliminate big mistakes in the basic clinical dosimetry. Repeated audit in the regional radiotherapy centers that had major deviation during the first audit exhibited improvement of their dosimetry. It is intended to broaden the method and to control also other beam parameters by means of a multi-purpose phantom. INTRODUCTION There are 75 irradiation units operating at the present time in the Czech Republic, delivering radiation therapy to cancer patients: 35 Co-60 units, 21 Cs-137 units, 15 linear accelerators and 4 betatrons (Figure 1). These irradiation units are placed in 34 regional radiotherapy centers, but there are o_nly 10 centers which are equipped conveniently (linear accelerator, CT, simulator, computer planning system, complete dosimetric equipment) to provide modern high quality radiotherapy. Structure of all irradiation units is shown in Figure 1. About 50% of all irradiation units are older than 10 years. In the Czech Republic (population approx. 10 million), the approximate 114 21" RHD Jasná pod Chopkom number of new oncology patients, who require radiotherapy, either curative or palliative, reaches about 15 000 per year. It is generally accepted that ±5% uncertainty in dose delivery to the irradiated volume is a safe limit causing no severe treatment consequences. Due to the complexity of procedures involved in radiotherapy, from the beam dosimetry, patient data acquisition and treatment planning, to the irradiation of the patient, the development and application of relevant quality assurance (QA) and quality control (QC) programs seems to be a key factor in reducing overall uncertainty associated with subsequent steps of the radiotherapy chain. According to a new law, which was recently implemented in the Czech Republic, each radiotherapy center has to undertake an independent quality audit every year. For performing audits, the State Office for Nuclear Safety (SONS), which is responsible for radiation protection in the Czech Republic, decided to establish an auditing group of experienced medical physicists working in the field, which is attached to the National Radiation Protection Institute (NRPI). It was decided by the auditing group that TLD postal audit combined with film dosimetry will alternate with "in situ" audit every two or three years. Therefore, a local TLD measuring network had to be established. The Czech local measuring center for TLD postal audit in the radiotherapy was established in the NRPI in Prague. Activities of NRPI reside mainly in providing of measurements and expert opinions for SONS's requirements. Consequently, NRPI's TLD audit is intended not only to improve clinical dosimetry in the radiotherapy centers, but also becomes an important source of information for state supervision in the radiotherapy. In addition, regular audits could help to get out of the use some bad irradiation units, and to contribute to change the bad structure of the regional radiotherapy centers. Figure 1: Structure of irradiation units in the Czech Republic • Cs-137 EHCo-60 ELinac(X) ELinac(X+E) • Betatron MATERIALS AND METHODS Pertinent methods for TLD audit were taken from European Measuring Center(1) and EROPAQ- EURAQA projects(2>3) in 1996 and were adjusted to conform to the Czech local conditions. For purposes of calibration and testing was used ^Co irradiation unit at Department of Oncology of 1. Medical Faculty - Charles University in Prague. First TLD audits were started in 1997. In the course of 1997 and 1998, a total of 135 2ľ RHD Jasná pod Chopkom 115 beams was checked: 45 60Co beams, 27 I37Cs beams, 29 X-rays beams and 34 electron beams. TLD system Lithium fluoride thermoluminescent virgin powder type MT-N (LiF: Mg, Ti — natural abundance, doped with magnesium and titanium) of Polish production was used for the irradiation and read with the Harshaw TLD reader - model 4000. Annealing of the powder was made by using a temperature cycle of 400°C/lhour and 100°C/2hours. The powder was encapsulated in waterproof capsules in portions, large enough to obtain 9-10 independent readings. Irradiated powder was dispensed into measuring containers (4 mm inner diameter, 20 mm inner length and 0.5 mm wall thickness) which were put onto reader's planchette. The following time-temperature reading cycle was used: Temperature 130°C (preheat) (rate) 10°C/s (max) 250°C Time (preheat) 8s (acquire) 20 s During all readouts the glow curves were recorded in order to eliminate possible errors due to the temperature shift of the reader. The TL-response of samples of irradiated powder to 2 Gy followed a Gaussian distribution with a mean value of 3503 nC and standard deviation of a single reading, a = 2.3%. Standard deviation of the mean (SD) for a single TLD capsule did not exceed 0.8%. (The experiment was made through reading of 20 capsules irradiated to 2 Gy on the same date, each capsule provided 9 samples). Organization of TLD postal audits Each participant (radiotherapy center) of the audit was provided with: • an instruction sheet describing the method of irradiation of TLD capsules • a data sheet to enter specifications of the therapy machine, dosimetry equipment, and details concerning TLD capsules' irradiation • the IAEA holder stand, in which TLD capsules were placed for irradiation • a group of TLD capsules Composition of dosemeters for TLD postal audit of one beam was the following: A/ Ancillary dosemeters Laboratory dosemeter • 1 capsule • placed in NRPI TLD laboratory • used for background measurement Transport dosemeter • 1 capsule 116 2 ľ' RHD Jasná pod Chopkom • sent to a hospital • used for background measurement Reference dosemeter • 3 capsules for irradiation to 2 Gy by a 60Co beam • used for determination of calibration factor B/ Audit dosemeters Dosemeters for beam calibration check • 3 capsules for irradiation • sent to a hospital Dosemeters for beam quality check • 4 capsules (= 2 pairs) for irradiation • sent to a hospital The participants were requested to irradiate the capsules in sequence to the absorbed dose of 2 Gy and to check the beam output with their dosimetry system before irradiation of TLDs. The irradiation in participating centers was done during a determined time window. At the same time, the NRPI irradiated one reference sample per center with 2 Gy from fi0Co. The irradiation of all capsules during a determined time period helped to control the fading and made better the ordering of the audits. The time between irradiation and reading usually didn't exceed three weeks. Irradiation Gamma and X-ray beams Capsules were irradiated in a water phantom using an IAEA holder stand. A/ For beam calibration check; capsules were inserted into the upper hole of the IAEA holder(4) at the depth of 5 cm. In case of a need of the depth of 10 cm (dependent on beam quality) the holder was lengthened by a distance indicator. The water level was adjusted precisely to the top of the holder and the axis of the beam was aligned with the holder axis. Field size of 10 cm x 10 cm and common SSD were set up. B/ For beam quality check, capsules were inserted into the both holes of the lengthened holder. After adjusting of the water level to the top of the holder, the upper capsule was at depth of 10 cm and the lower at depth of 20 cm. Field size of 10 cm x 10 cm and SSD = 100 cm were set up. Electron beams For beam calibration check, capsules were inserted into the hole of the special (4) IAEA holder to be irradiated at the depth of dmax. Correct positioning of the capsules at the depth of dmax was made setting a corresponding number of spacers. The water level was adjusted precisely to the top of the holder and the axis of the beam was aligned with the central holder axis. Field size of 10 cm x 10 cm and common SSD were set up. 21" RHD Jasná pod Chopkom \ \ 7 Dose calculation The absorbed dose to water, D (Gy), at the location of TLD was calculated from the TL signal, R, registered by the reader using the following formula D = R Kcal Klin Kfad Ken where: R is the TLD reading normalized to the mass of aliquot of the powder, Ktaj (Gy/nC) is the calibration factor determined for 2 Gy from 60Co beam, Kiin is the dose linearity correction determined on the basis of experimentally determined linearity function, Kaj is the fading correction determined on the basis of experimentally measured fading function, K«n is the X-ray energy response correction determined by comparing the TLD response to the same dose to water in a high-energy X-ray beam and 60Co beam, under reference conditions. Reporting and analysis of deviations The TLD measured values were compared to the values stated by physicist, who irradiated the TLD capsules. For each capsule the mean reading value and standard deviation were determined. The average reading value of three capsules (evaluated absorbed dose) was calculated. A/ For beam calibration check, a deviation between measured (TLD) and stated dose is reported: AD = (DTLIJDS- 1)100% B/ For beam quality check, a deviation between measured (TLD) and stated QI is reported: A&= (QITLD/QL-1)100% Analysis of deviations was based on values of A: /A/Š3% means acceptance level, beam complies requirements of quality assurance program 3% < lál £6% means minor deviation, investigating causes and/or repeating of TLD audit is necessary, means major deviation, "in situ" audit in details is necessary Quality control of TLD system The accuracy of TLD measurements has been verified regularly by means of different intercomparisons - mainly comparison of results measured by NRPI, IAEA 118 2ľ RHD Jasná pod Chopkom and EROPAQ/EURAQA Measuring Center (University Hospital Gasthuisberg, Leuven) under the same conditions. Most of the Czech radiotherapy centers have been checked also by in-situ audit, so results measured by TLD and ionizing chamber are compared regularly(5). All these intercomparison results showed very good agreement or at least compatibility. The 60Co unit that is used for calibration is regularly checked by ionizing chamber and by intercomparison measurements with the Czech Secondary Standard Dosimetry Laboratory. RESULTS AND DISCUSSION The results of the first course of TLD audit, which were done in the Czech Republic from 1997, are summarized in Figure 2. Figure 2 shows distribution of deviations AD for beam calibration check during the first TLD audit. In the course of 1997 and first half of 1998, a total of 110 beams was checked: 33 60Co beams, 19 137Cs beams, 26 X-ray beams and 32 electron beams. Most of the checked beams comply with the acceptance level, but there were seven beams with major deviations (2 137Cs beams, 2 60Co beams, 1 X-ray beam of betatron and 2 electron beams of betatron). The deviations were very carefully analyzed, and it was reported by medical physicists from the departments that they had made a set-up mistake. However, at most of those hospitals with major deviations, an in situ audit in details was made soon after TLD audit. There were found discrepancies of clinical dosimetry but also bad technical state of some of the irradiation units. In 1998, second TLD audit was started - there were checked preferably beams with major and minor deviations that were found during the first TLD audit, also beams of old irradiation units and beams of radiotherapy centers with inconvenient equipment. Results are shown in Figure 3. A total of 25 beams was checked: 12 ^Co beams, 8 I37Cs beams, 3 X-rays beams and 2 electron beams. It is evident that the results exhibit better distribution of deviations. No major deviations were found. The TLD postal audit seems to be an efficient way to eliminate big mistakes in basic clinical dosimetry. CONCLUSIONS A new law for radiation protection in the Czech Republic has enforced the necessity of TLD network i radiotherapy. The TLD postal audit contributes to improvement of clinical dosimetry because its results are an important part of data for state supervision in the field of radiotherapy. The results can support the desirable change of the structure of the irradiation units in the Czech Republic. It is desirable to supplement some of the old 60Co units and betatrons by new modern linear accelerators, and to get out most of the 137Cs units from the clinical practice. The necessity and usefulness of TLD audit were sustained. The regular control can press regional radiotherapy centers to improve their dosimetry. The regional centers, which exhibited major deviations in the first course of TLD audit, had better results of the second audit. The control made by TLD will be broadened in the future. It is intended to start TLD audit by means of a multi-purpose phantom in 1999, the TLD audit combined with a film dosimetry and a multi-purpose phantom could partly supplement an in-situ audit in the radiotherapy department, because basic beam data can be obtained from these 2 ŕ RHD Jasná pod Chopkom 119 Figure 2 : Results of l'TLD audit - distribution of A> - beam calibration check DCs-137 EICo-60 DLinac(X) BLinac(E) • Betatron (X) • Betatron (E) Figure 3: Results of 2nd TLD audit - distribution of AD - beam calibration check 15 12- r- L L -. • f 9 S W IT - í. _ - -• mm 3 1J 1 _l ML_ 1 rzL - — -3 0 3 >9 AD[%] DCs-137 SCo-60 DLinac(X) QLinac(E) • Betatron (X) • Betatron (E) 120 2ľRHDJasnápodChopkom measurements. In our country we propose that postal TLD audits will be performed every 1 or 2 years for each clinically used beam. In situ audit in details should be made within a period of 3 years. In the case that a department will exhibit permanently major deviation then in situ audit can be requested more frequently. REFERENCES 1. Derreumaux, S., Chavaudra, J., Bridier, A. A European Quality Assurance Network for Radiotherapy: Dose Measurement Procedure. Phys. Med. Biol. 40,1191-1208 (1995) 2. Izewska, J., Novotný, J., Gwiazdowska, B. Quality Assurance Network in Central Europe: External Audit on Output Calibration for Photon Beams. Proceedings of EROPAQ and EURAQA Meeting, Warsaw, 9-10 June 1995. U.H. Gasthuisberg, Leuven, Belgium 3. Izewska, J., Dutreix, A. Guidelines for Organising TLD Audits of the Beam Calibration in Radiotherapy Departments at the National Level. 4th Progress Report, 1997, U.H. Gasthuisberg, Leuven, Belgium 4. Izewska, J. Guidelines for Organising TLD Audits of the Beam Calibration in the Radiotherapy Departments at the National Level IAEA, (1997) 5. Kroutilikova,D., Zackova, H., Novotný, J. TLD Postal Audit in the Czech Republic. Proceedings of EROPAQ and EURAQA Meeting, Prague, 27-28 June 1997. U.H. Gasthuisberg, Leuven, Belgium 2ŕRHDJasnápodChopkom SK98K0355 121 RADIATION PROTECTION PROBLEMS BY THE OPERATION OF THE CYCLOTRON FACILITY Matej Ďurčik and Denisa Nikodémova Institute of Preventive and Clinical Medicine Limbová 14, 833 01 Bratislava, Slovak Republic Introduction Accelerator operation can create substantial hazard. Effective safety programs should therefore, be implemented to reduce, or even eliminate, hazards to the workers, to the public and to the environment. Radiation safety system of cyclotron consists of passive shielding, active ventilation systems, monitoring plan and safety lock systems. In our contribution are described all proposed radiation safety systems in Slovak Cyclotron Center except of passive shielding of high radiation areas. The Cyclotron Center will consist of two cyclotrons. First - cyclotron DC-72 with maximal energy of 72 MeV for protons designed for making experiments, for teaching process, for radioisotope production as 123I and for neutron and proton therapy. Second - compact cyclotron with maximal proton energy of 18 MeV will be used for radioisotopes production for medical diagnosis as 18F (FDG), 8!Rb/81Rr generator.[l] The dispersal of the radioactive gases produced by neutron activation is one of the most important problems to be considered in evaluating the hazards to the workers and the local population. Neutron may be produced in the cyclotron vault (because of beam losses inside the cyclotron, particularly at the extraction points and on the beam transport system) and in the target room (by the beam striking the targets). The interaction of these neutrons with components of airs mainly produces following radioisotopes: 13N, 15O, I6N,37 S, ""Cl and ^Ar in cyclotrons with energies up to 40 MeV for protons. During the irradiation also radioisotopes as 18F, 81Kr and I23I could escape from target system into the air.[2] Radiation protection methods Radiation protection of workers has very important role in decreasing of radiation hazard of people. The working area of Cyclotron Center is divided into three areas depending on the actual radiation level: • High radiation areas are places where radiation hazard could be higher than 50 mSv/a or 25 uSv/h. They involve cyclotron, target, dual beam and machine part exposition vaults, experimental and neutron hall. • Radiation areas. The radiation hazard could be higher than 15 mSv/a or 7,5 uSv/h. They involve all rooms where workers use the radioactive materials and where irradiated parts of accelerator equipment are located. • The other rooms in the vicinity of cyclotrons. People that work in the radiation area must wear two types of personal dosimeters (photoluminiscent glass dosimeter and pocket dosimeter with immediate reading possibility). The determination of the dose equivalents for personnel working in 122 2 ľ RHD Jasná pod Chopkom the cyclotron environment where photons and neutrons of wide range energies are simultaneously produced is a difficult task. Modified albedo dosimeter was introduced as a sensitive personal (n, gamma) dosimeter, which together with Bonner spheres would meet all requirements concerning accuracy and sensitivity. [3] Contamination monitors, dose rate telescope monitors and body contamination monitor are located in radiation areas. Monitoring plan consists of following activities: • occasional checking of the monitoring apparatuses and their calibration once per year, • surface contamination measuring in the labs and vaults, wiping the reference points and measuring alpha and beta contamination once per week, • measuring of dose rate in the labs once a week and the dose rates determination in the high radiation area before work, • dose rate measuring of radioactive waste and irradiated parts, • dose rate and surface contamination measuring of the produced radioisotopes and their preparation for transport, • aerosol paper filters and iodine charcoal filters changing and checking the beta air monitors twice per week (on Tuesday and on Thursday), • occasional dose rate measurement around the storage places and the containers with radioactive materials, • personal dosimetry and dosimetry of visitors. Ventilation must be organized separately for cyclotron vaults, therapy and diagnostics rooms. There must be two different ventilation systems in Cyclotron Center. One of them is used for ventilation and climatisation of no radiation area (control room and power supply rooms). The second is used for ventilation of radiation areas and is devided in two lines, first line for vaults and second line for chemical boxes and hot cells. Ventilation rate is constant and equal to 6 air exchanges per hour in radiation areas. Through ventilation the radioactive gases are released to atmosphere and radiation hazard of workers is reduced because the probability of escaping this gases created in vaults into the working areas is minimized. The ventilation creates underpressure about 50 Pa in vaults and hot cells. There should be problems with release activities to the environment and these activities must be monitored and limited. The limits for different groups of released radioisotopes should be establish and their influence on environment has been discussed. Safety lock and limitation are very important for radiation safety of workers in nuclear facility and environment. The lock system consists of the dose rate monitors placed in the vaults, indicators placed outside near the entrance doors into the vaults and of the control system. When the cyclotron is switched on or dose rates of gamma radiation are higher as 100 uSv/h in the vaults the doors are locked and could be opened only in a case of accident after switching off the cyclotron by a special key and person. The safety department and deputy health officer have determined limits and rules for radiation protection of environment and working places. When the cyclotron releases 80% of the limit for a week during a few days or exceeds the limit, the 21"RHD Jasná pod Chopkom 123 cyclotron must be switched off until next week. Supervisor must necessarily search for reason why was released high activity to environment. The released activities from similar cyclotrons per year are usually about 10% of the limits [4,5] and the results show that no significant radiological risk will be produced by the release of the radioactive gases into the external environment. CH Isokinetic sampler F -i PF D D n CF DATA& w CONTROL Treatment Cyclotron Isotope Lab & Hot cells Experimental STATION rooms vaults production QA & boxes rooms vaults Figure 1 Block diagram of the proposed monitoring and safety system in the KAZ and central data acquisition. CH - chimney, D - dose rate meters, PF - aerosol filters, CF - carbon iodine filters, FM - flow rate meter, GM - beta gas monitor, Int - multi logger and F - output filters. Measurement techniques Adequate monitors determine the quality of radiation monitoring. The choice of them depends on the type of radioisotopes induced by activation during the cyclotrons operation. The monitoring system in Cyclotron center consists of measuring: • alpha and beta aerosol activities in the radioisotope production laboratory which could escape from transporting tubes and hot cells, • iodine activities in the radioisotope production laboratory, • gamma dose rate in the radioisotope production laboratory, • beta air activities release to the atmosphere through chimney, • alpha and beta aerosol activities release to the atmosphere through chimney, • iodine activities release to the atmosphere through chimney. The aerosols are sampled on paper filters and the collected aerosol activities are measured by alpha-beta proportional counter. The radioactive iodine isotopes are collected on charcoal filters that are measured by gamma spectrometry. The Central 124 2ľ RHD Jasná pod Chopkom safety department is responsible for measuring these filters and determination of their activities. For aerosols and iodine activities determination in the labs and chimney air are used the same types of filters and collectors. The Figure 1 shows a block diagram of ventilation monitoring devices for high radiation areas of Cyclotron Center. [6] The beta active and noble gases in air released through the chimney are measured by monitor LB-5310 with gas-flow proportional counter and 86 liters or 4,6 liters measuring chambers Conclusion In accordance with the Basic Safety Standards No.115 and ICRP recommendations No.60, for proposed and continuing practices at the accelerator facility, the following general principles have to be fulfilled: • practices should produce sufficient benefit to offset the radiation detriment they cause (justification) • the magnitude of the individual doses should be kept as low as achievable (optimization of protection) • individual exposures are subject to dose limits and some control of risk from potential exposures (dose and risk limits). As the nature of cyclotron radiation is such, that while routine doses are small, effective safety programs should be implemented and preventive measures emphasized to assure the reduction of the hazards to the public, workers and patients at early stages of cyclotron facility design. References [1] Správa zo seminára o stave prác na kontrakte 85-015-70700 pre CyLab. Moskva 1998. [2] A.H.Sulivan: A Guide to radiation and radioactivity levels near high-energy particle accelerators. NTP, ISBN 1-870965-18-3, England 1992. [3] D.Nikodemová, J.Staňo a O.Szollos: Možnostii vzužitia albedo dozimetra pre sledovanie osobných dávok neutrónov v cyklotrónovom laboratóriu SR, Zborní abstraktov NUCLEONIKA, Praha 1998. [4] C.Birattari, A.Ferrari, C.J.Paarnell and M.Silari: Radiation Protection Dosimetry Vol.19, No.3,pp.l83-186,1987. [5] W.Koelzer: Jahresbericht 1996 der Hauptabteilung Sicherheit. Forschungszentrum Karlsruhe, ISSN 0947-8620,1997. [6] P.Nemecek: Exhaust air monitoring systems in PET centers with Cyclotron, hot cells and labs. Berthold, 1994. 21 RHD Jasná pod Chopkom SK98K0356 125 VERIFICATION OF NUCLEAR MEDICINE RADIONUCLIDE CALIBRATORS IN SR. Anton Švec Slovak Institute of Metrology, Karloveská 63, 842 55 Bratislava The Slovak Institute of Metrology started with nuclear medicine radionuclide calibrators verification in 1997. There are 11 nuclear medicine departments operating up to 18 instruments for the assay of the radionuclides applied to patients "in vivo". The Metrological Law requires verification of these instruments annually. For the purpose, CURIEMENTOR 2 measuring device of PTW Freiburg production has been used as a travelling standard. The instrument was purchased and put into operation at the end of 1996 and checked for elementary performance for few months. In April 1997 basic calibration was realized at Czech Metrological Insitute - Inspectorate for Ionizing Radiation with some selected radionuclides: 99mTc, I31I, 137Cs and 226Ra. Calibration conditions for the first three nuclides were 5 ml of water solution (or its gelic equivalent in the case of 137Cs) in a standard penicilíne ampoule (of ÚJV Rež production) while 226Ra was used in the "needle" form (EP standard according to STN 40 4401). Calibration constants obtained were introduced in the instrument memory and used during verification actioa The difference between calibration coefficients given originally by the producer and new ones was almost negligible for I31I and 137Cs, however, for 99mTc the difference was unexpectedly high (more than 10%). This fact was regarded so important that no measurement of nuclides other than calibrated ones was offered to nuclear medicine laboratories. During May-July 1997 15 radionuclide calibrators were verified, among them 8 of the same type like the standard one and 7 of various other types (Mediae Dose Calibrator, Picker Isotope Calibrator, Robotron M27013, TESLA NRR601 and Victoreen). It was first such an action in the independent Slovak Republic [1]. Unlike other more developed laboratories, Slovak Institute of Metrology has no possibilities to prepare calibrated standard radioactive sources so only "in situ" prepared solutions vere used. The verification conditions were kept as close as possible to the calibration conditions although not identical: the ampoules were used of the type typical for each workplace and the user's ionisation chamber was left in its usual position in shielding while reference chamber was unshielded. The basic idea for doing this was to transfer the reference chamber calibration to typical working conditions in the user's laboratory without any additional corrections. The results were published [2] and were regarded very startling because 60% of instruments did not fulfil the condition of the 10% difference between verified and reference readings. No recalibration was done during this first action. Besides the verification of instrument calibrations for 99mTc and 131I radionuclides, also data and readings of 137Cs check sources (when used) were obtained and compared with both 137Cs standard measurements and measurements of the check sources by the travelling standard instrument. This cross-check proved that check sources, although provided with a certificate, cannot be regarded as activity standards except the half-time of the radionuclide used; i.e. their regular use gives information about the instrument stability but not its accuracy. Moreover, older instruments have their selectors of radionuclides arranged in such a way that a correct measurement of a nuclide does not 126 2 ľ RHD Jasná pod Chopkom necessary mean that other nuclides would be measured correctly too. Therefore check measurements with these instruments should be performed and registered at all used radionuclide ranges. This is not the case of CURIEMENTOR 2 which has calibration coefficients saved as a multipliers in its memory so a correct reading at one selected range can be regarded as a proof that other readings will be correct too unless calibration coefficients are changed. In October 1997 an opportunity to perform an international comparison in PTB Braunschweig was taken and utilized for an energy calibration curve construction [3]. The curve was constructed from measurements of 18 long-lived radionuclide standards and verified by comparison of measured and calculated data for some polyenergetic radionuclides like Cs, 133Ba and l52Eu as well as previous calibration. Calibration conditions were slightly different with former ones namely the solution volume was 2 ml only and injection ampoules were used. Yet the agreement with the first calibration within 1% was acchieved for 99mTc, 131I and 137Cs nuclides and the overall uncertainity 1,5% using calibration curve was declared. Based on the calibration curve, all available calibration coefficients of the standard instrument CURIEMENTOR 2 were calculated and compared with original ones. Most important differencies were found at 57Co (-11,3%), 67Ga (-7,8%), WmTc (- 9,2%), 113mIn (-9,0%), 125I (+22,7%) and 201Tl (-12,6%) which are all frequently used in nuclear medicine praxis. This fact stresses the importance of primary calibration and verification even of new instruments introduced into operation, not regarding producer's guarantee, as well as the necessity to verify instruments after their repair. The standard instrument is regularly checked with four 137Cs check sources in the range from 3,7 MBq to 37 GBq. This check serves as a proof of its linearity up to high activities. Another check with 137Cs and 226Ra standard sources proves that the basic calibration has not changed. All this checking is performed monthly and before and after each transport of this travelling standard. One-and-half year of its operation has shown that its sensitivity is gradually degraded by about -1% a year. The most probably cause seems to be a leak in the pressurised chamber which cannot otherwise be observed. CMI has similar knowledge about such a behavior of these instruments. In April this year CMI send us a 131I sample for comparison. The sample was a standard one (5 ml in PNC ampoule) so no corrections except for radioactive decay and background were necessary. The result of this comparison -1,2% was found in good agreement with all previous observations. In July 1998 another verification of nuclear medicine laboratory instruments was made in Slovak Republic. The number and choice of the instruments was the same like last year (just one old Victoreen instrument has been replaced by a Mediae Dose Calibrator) so direct comparison of results is possible. The situation has greatly improved also due to several instruments were recalibrated "in situ". Although six radionuclides were checked (99mTc, 131I, 20ITl, 67Ga, 123I and 51Cr) only two of 15 instruments (both of Mediae Dose Calibrator type) did not pass the verification. Such a satisfactory situation could be at least preserved if proper attention to the instruments technical conditions is paid. SMÚ has done its best to reach an internationally acceptable level with its equipment for activity measurements which create the National Standard of Activity. The instrumentation is to be widened by a stationary ionization chamber which replaces CURIEMENTOR 2 in its standard function and is expected to be more long-term reliable. Both chambers are going to be calibrated abroad and will create a tandem with low uncertainty and high reliability serving not only nuclear medicine laboratories [4]. 2 ľ1RHD Jasná pod Chopkom 127 Literature: [1] Šuráň J., Jasanovský P., Dryák P.: Verification of radionuclide calibrators used in nuclear medicine laboratories in the Czech and Slovak Republics. Proc.rV.Conf.Rad.Prot., Orlando, U.S.A. 1993. [2] Švec A.: Overenie urěených meradiel aktivity rádiofarmak (Verification of radionuclide calibrators). Metrológia a skúšobníctvo 3 (1998), in press. [3] Schrader H., WeiB H.M.: Calibration of Radionuclide Calibrators. Int.J.Nucl.Med.Biol. 10 (1983), 2/3,121-124. [4] Rytz A.: The International Reference System for Activity Measurements of y-Ray Emitting Nuclides. IntJ.Appl.Radiat.Isot. 34 (1983), 8,1047-1056. 128 SK98K0357 2ľ RHD Jasná pod Chopkom RECENT STAGE OF THE EVALUATION OF MEDICAL EXPOSURES IN THE CZECH REPUBLIC AND DIAGNOSTIC RADIOPHARMACEUTICAL DOSE ESTIMATE TO THE CZECH POPULATION Karla Petrová1*, Václav HuSák2* 1) State Office for Nuclear Safety, Senovážné nám.9, 100 00 Praha 2) University Hospital Palackého, Nuclear Medicine Dept., I.P.Pavlova 6, Olomouc Introduction The attention of regulatory authorities is recently much strongly focused to the evaluation and regulation of medical exposure. State Office for Nuclear Safety ( SONS) co-operate on this field with the General Health Insurance comp., which provide it with very valuable data, which enable us to sort examined persons going through the individual procedure by the age and sex (see paper Petrová ,Prouza ). There is also possibility to get the information about the administered activity in the case of nuclear medicine procedures. Methods, analysis, results Based on this data the frequencies of diagnostic procedures, average administered activity of radiopharmaceuticals ( RAF) and collective effective dose arising from the practice of nuclear medicine in the Czech Republic in one year period of 1995 and 1996 years, were estimated. The General Health Insurance Comp., covering' 75% of the Czech population provided the necessary data in the frame of the project sponsored by the Czech Grant Agency. The values of conversion factors mSv/MBq for most considered RAF were supplied by the courtesy of M.G.Stabin ( Oak Ridge,USA). Taking into account the age distribution of patients and age dependent conversion factors, the mean annual effective dose per caput and per exam in CR amount to 0,09mSv and 4,8mSv, respectively. These figures as well as 21 administrations per thousand of population are lower than those in USA and Germany, but higher than in most European countries. The relative distribution of examination in nuclear medicine and their contribution to the collective effective dose shows Fig.l. Fig.l: The relative distribution of examination in nuclear medicine and their contribution to the collective effective dose 2 ľ' RHD Jasná pod Chopkom 129 "Tc is by far most common radionuclide, being used in 98% of administrations and contributing nearly 84% of the collective effective dose. Based on the determined average values of administered activity during this work, the guidance levels for nuclear medicine procedures were established and implemented into a new Czech legislation ( Regulation No. 184/1997). In comparison with the similar survey performed in our republic in 1987 there is 40% increase in the collective effective dose ( 860 man Sv) increase in average activity of RAF administered ( e.g. from 103MBq to 250MBq "Tc - MAA, from 556MBq to 730MBq 99mTc- phosphates etc.) as well as increase in the annual number of some procedures ( e.g. bone scintigraphy by factor of 4). Fig.2 10,00 I Literatura: Í.Petrová K., Prouza Z.; The National Central Registries of Occupational and Medical Exposure in the Czech Republic, IRPA 9 Conference Proceedings, Vienna, Austria, vol.4, 682-684 pp (1996) 2.Petrová K., Husák V., Masopust J., Prouza Z. .; Hodnocení ozáření pacientů při použití zdrojů ionizujícího záření v nukleární medicíně v ČR, Sborník abstrakt XXXHI.Dnů nukleární medicíny, Hradec Králové, (1996) 130 SK98K0358 21"RHDJasnápodChopkom INDIVIDUAL DOSIMETRY IN HIGH ENERGY RADIATION FIELDS František Spurný Oddělení dozimetrie zářeni, Ústav jaderné fyziky A V ČR, Na Truhlářce 39/64, 18086 Praha 8 Individual dosimetry in high energy radiation fields is of interest at high energy particle accelerators, in cosmic rays radiation fields at altitudes above ~ 8 km and, perhaps, also in connection with proposed accelerator driven transmutation technologies Generally two basic methods are used to determine dosimetric characteristics in a radiation field: The first one is based in the first step on the determination of the absorbed dose, independently on the type or energy of particle transferring the energy. For radiation protection the second step is needed in which full LET spectrum of energy transfer processes is determined and dose equivalent and/or the effective dose are calculated. The second method is based on the separate determination of the radiation protection quantities for each component of radiation field studied, total dose equivalent is obtained as a sum of individual particle contributions. Typical example of the first method in mixed radiation fields of usual energies represents for field monitoring the use of tissue equivalent proportional counters (TEPC) based on microdosimetry principles [2]. For the second method is for same fields typical the use of a compensated GM counter for photon component, resp. a moderated „remmeter" for neutrons. As far as individual dosimetry is concerned the situation is not ideal neither for lower energy (S 20 MeV) mixed radiation fields, particularly for the individual neutron dosimetry [3]. The situation is still more complex in the case of individual dosimetry for high energy (> 20 MeV) radiation fields. These fields are typical by large diversity in the particle types, their energies and, also, with very complex and large spectra of linear energy transfer (LET). At the same time, there is not yet any adequate individual dosimeter based on the first method mentioned above. The energy transfer processes are different from low energy region in many aspects: • particle transport is controlled by a cascade processes (hadron, or electromagnetic); • the identity of primary particles is during these processes modified; • the ranges of particles are high, even greater than the human body; • the importance of nuclear interaction to particle's transport and their energy transfer is much higher. 2ŕ RHD Jasná pod Chopkom 131 We have tried to verify the usefulness of some existing individual dosemeters in some high energy fields: in the on Earth's reference fields at CERN, Geneve [12], and at the JINR at Dubna [13], and on aircraft board. The reading of different type of individual dosemeters (see Table 1) were directly compared with reference data obtained generally with a TEPC instrument. The readings of tested dosemeters were expressed by means of Hp(10) of the reference radiation, i.e. 60Co photons for low LET component of a high energy field, resp. AmBe neutrons for high LET component. The results obtained expressed relatively to the TEPC data are presented in the Tables 2 and 3. Table 1: Individual dosemeters tested in high energy radiation fields Dosemeter Component of radiation Photographic film PS1 [4] Al-P TL glass [51 7LiF in albedo dosemeter PGP DIN [4] Electronic, Si-based, CAFP-JINR, Dubna [6] low LET Electronic, Si-based, DMC-France [6] Electronic, GM-based, D222-Czech Rep. [6] Nuclear emulsion NTA [4] 6LiF in albedo dosemeter PGP DIN [4] Track etch detector (TED) - PADC (CR39) [7] TED in contact with fissile radiators [5] high LET Bubble detectors, BDND (100) ľ81 (neutrons) Superheated drop detectors (100) [9] Superheated drop detectors (6000) [9] Table 2: Relative responses of individual dosemeters to low LET component in some high energy radiation fields Dosemeter Relative response at JINR hard CERN top concrete board aircraft Photographic film - 0.810.1 - Al-P TL glass 1.2±0.2 1.2±0.2 0.9±0.2 7LiF in albedo - 1.5±0.2 - Electronic (Si-CAFP) 0.6±0.1 0.9±0.1 0.7±0.1 Electronic (SÍ-DMC90) 0.5±0.1 0.8±0.1 - Electronic (GM-D222) - 1.6±0.2 1.010.1 As far as the high LET component detectors are concerned, the situation is less favourable. Albedo as well as TED (PADC) are relatively low sensitive, nuclear emulsions and TED with fissile radiators overestimate the exposure level. 132 21" RHD Jasná pod Chopkom Table 3: Relative responses of individual dosemeters to high LET component in some high energy radiation fields Dosemeter Relative response at JINRhard CERN top concrete board aircraft Nuclear emulsions - 3.7+0.9 - 6LiF in albedo - 0.10±0.02 - TED-PADC 0.37±0.05 0.35+0.06 - TED with fissile rad. 3.5±0.5 2.9+0.5 - BDND (100) 0.7±0.1 0.610.1 0.5+0.1 SDD (100) - 0.6±0.1 0.6±0.1 SDD (6000) - 1.1+0.2 1.0±0.2 One can see there that the relative responses of low LET component dosemeters are generally close to 1.0, the underestimation observed for electronic dosemeters in JINR - hard field can be attributed to the saturation due to very fin beam structure and corresponding very high instantaneous dose rates. The usefulness of an individual dosemeter depends not only on its relative response but also on its absolute sensitivity, i.e. the threshold value in dose equivalent which can be established with sufficient statistical reliability. These threshold values are presented in Table 4. Table 4: Threshold values of Hp(10) for individual dosemeters tested Low LET component High LET component Threshold p.Sv Threshold, (xSv Dosemeter high energy Dosemeter Reference reference radiation radiation radiation nuclear photographic film 60 200 60 emulsion Al-P TL glass 10 TED-PADC 100 300 Electronic Si-CAFP 0.1 TED-fission 300 100 Electronic Si-DMC 10 BDND (100) 10 20 Electronic GM-D222 10 SDD (100) 10 20 SDD (6000) 30 30 To discuss them it should be reminded that the detection threshold of an individual dosemeter should be, taking into account ICRP 60 recommendations, about 80 (iSv for 1 month period of using [10]. From that point of view, all low LET component individual dosemeters tested fulfil this requirement. As far as high LET component (neutron) individual dosemeters are concerned, their effective thresholds for high energy radiation (i.e. corrected for relative response) are favourable for both types of bubble detectors and nuclear emulsions, still acceptable for TED's in contact with fissile radiators. 2 ŕRHD Jasná pod Chopkom 133 To conclude it should be emphasided that also some practical aspects have to be taken into account to choose an individual dosemeter for the use in a high energy radiation field. For example: • there are some problems with the routine use of bubble detectors due to the sensitivity to temperature and mechanical shocks; • the large scale distribution of fissile radiators could arise some problems; • nuclear emulsions can be used only when necessary, rather complicated, laboratory equipment and procedures are available; • last but not least, the cost should be also taken into account. Finally, it should be emphasised that all individual dosemeters tested are based on the second method as defined in the introduction, inevitably more complicated to interprete in an unknown radiation field. There is therefore stil a great challenge to solve the most of still existing problems: to develop for large scale use an individual dosemeter based on the principle of microdosimetric tissue equivalent proportional counter [11]. REFERENCES [1] SWANSON, W.P. - THOMAS, R.H.: „Dosimetry for Radiological Protection at High Energy Particle Accelerators." in „The Dosimetry of Ionizing Radiation"; eds.: Kase, K.R. et al., Academic Press (San Diego), 1992, vol. 3, p. 1-162 [2] „Design, Construction and Use of Tissue Equivalent Proportional Counters." EURADOS Report, special volume, Radiat. Prot. Dosim. 1995, 61, No. 4; eds.: MenzeL, H.G. et al. [3] „Neutron Dosimetry." Proc. of the 8th Symp., Paris 1995; special volume, Radiat. Prot. Dosim., 1997, 70, No. 1-4; eds.: Menzel, H.G. et al. [4] „La dosimétrie individuelle au laboratoire ďexploitation dosimétrique de Fontenay-aux-Roses." LED/DPHD/IPSN, CEA Fontenay-aux-Roses, 1990 [5] TROUSIL, J. - PROUZA, Z. - STUDENÁ, J.: „Basic Dosimétrie Characteristics of the Czechoslovak Thermo luminescent and Neutron Dosemeters." Kernenergie, 1984,27, p. 246-254 [6] SPURNÝ, F.: „Experimental Approach to the Exposure of Aircrew to Cosmic Radiation." Radiat. Prot. Dosim. 1997, 70, p. 353-356 [7] TUREK, K. - SPURNÝ, F. - ALBERTS, W.G.: „On the Optimisation of the Etching of CR39 as Fast Neutron Dosemeter." Nucl. Tracks Radiat. Measur. 1993, 21, p. 299-302 [8] ING, H. - McLEAN, I. - NOULTY, R. - MORTIMER, A.: ,3ubble Detectors an the Assessment of Biological Risk from Space Radiation." Radiat. Prot. Dosim., 1996, 65 (1/4), p. 421-424 [9] D'ERRICO, F. - ALBERTS, W.G.: „Superheated Drop (Bubble) Detectors and their Compliance with ICRP 60." Radiat. Prot. Dosim., 1994, 54 (3/4), p. 357-360 [10] PORTAL, G. - DIETZE, G.: „Implications of new ICRP and ICRU Recommendations for Neutron Dosimetry." Radiat. Prot. Dosim., 1992. 44(1/4). p. 165-170 [11]BARTHE, J. - BORDY, J.-M., LAHAYE, T.: „Electronic Neutron Dosemeters: History and State of Art." Radiat. Prot. Dosim., 1997, 70 (1/4), p. 59-66 134 2ľ' RHD Jasná pod Chopkom [12] HÓFERT, M. - STEVENSON, G. R.: „The CERN-EC High Energy Reference Fields Facility." presented at the 8th Inter. Conf. on Radiation Shielding, Arlington, Texas, April 1994 [13] ALEINIKOV, V.E. - BAMBLEVSKIJ, V.P. - KOMOCHKOV, MM. - KRYLOV, A.R. - MOKROV, J.V. - TIMOSHENKO, G. N.: „Reference Neutron Fields for Metrology of Radiation Monitoring." Radiat. Prot. Dosim., 1994,54, p. 57-60 ACKNOWLEDGEMENTS: Author is obliged to the Grant Agency of the Academy of Sciences of CR for the financial support through the contracts No. 335402 and 3048606 21" RUD Jasná pod Chopkom S K98 K0359 135 MONITORING OF THE INTERNAL CONTAMINATION IN THE CZECH REPUBLIC. SURVEY FOR THE NEEDS OF EURADOS GROUP Irena Češpírová, Irena Malátová National Radiation Protection Institute, Šrobárova 48,100 00 Praha 10 The European Radiation Dosimetry Group (EURADOS) is an international scientific organisation founded in 1981 by members from laboratories in Europe working in the field of radiation measurement and dosimetry. It is concerned with advancing scientific understanding and technical development of the dosimetry of ionising radiation in the fields of radiation protection, radiobiology, radiation therapy and medical diagnosis by stimulating the collaboration between European laboratories, especially between those in the European Union. It was decided to compile a complementary database, including information on materials (radionuclides, physico-chemical form etc.), field of application (industry, medicine, research), scale (numbers of persons, activities) etc. It was recognised that a simple method of collecting the information was needed, and therefore a questionnaire was developed for members to distribute to relevant organisations in each country. A test of the draft questionnaire (in Germany) provided important results which were incorporated in a revised version. These related to the definition of "exposed" and "significantly exposed" persons, now defined as those monitored and those in whom activity has been detected. Two countries outside of European Union participate in some of EURADOS activities. The Czech Republic (NRPI) has been involved into 2nd and 3rd Intercomparison of doses from internal contamination based on data from whole body counting and bioassay. NRPI collected ako data for the database mentioned above. Information collected by NRPI and submitted to EURADOS group are presented in the following tables. The work was partially supported by the Grant No 4965-3 of the Internal Grant agency of the Ministry of Health of the Czech Republic. 136 21" RHD Jasná pod Chopkom Tabl. Field codes Field codes | Field of activity Nuclear industry 1 Mining 2 Milling 3 Reactor operation 4 Waste management and disposal 5 Others (specify) - maintenance NPP 5A - research reactor Non-nuclear activities 6 Industry - production of radio farmaceutic - use of radionuclides in gauges etc. - production of sources for smoke detectors - production of radionuclides - production of labelled organic compounds 6A - construction of shielding / containers from depleted uranium 7 Medicine - hospitals (diagnostic, therapeutic and palliative use) 7A - radon Spa 8 Research 9 Waste management and disposal 10 Other (specify) - Metrology Tab2A. Number of workers in nuclear industry Field code 1 2 3 4 5 5A Total persons involved 500 100 2396 28 67 214 Routinely monitored 500 100 155 28 67 214 a Exposed persons ) "") •") 62 1 15 2 Tab2B. Number of workers in non-nuclear activities Field code 6 6A 7 7A 8 9 10 Total persons involved 60 42 81 84 77 19 25 Routinely monitored 56 42 76 64 77 19 0 Exposed personsa) 6 4 °) a) Exposed person are those for whom internal contamination has been detected b) Doses calculated from personal dosimeters, measuring external irradiation, EEC and long lived aerosols c) Doses calculated from measurements of EEC Tab 3 Routine monitoring methods Method Description of method WBC whole-body measurements Thyroid measurements of thyroid Urine measurements of urine Air air concentration monitoring Tab 4. Monitoring results Field (code) 3 3 3 3 3 3 3 5 5 5 5 5 Nuclide llumAg ""Co i8Co ^Mn 95Nb JH "'I MMn 58Co wCo "Fe s>Nb Compound (found or assumed) HTO Assumed Pathway (ing./inhal.) inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. AMAD [umj assumed 1 1 1 1 1 1 1 1 1 1 Monitoring method WBC WBC WBC WBC WBC Urine Thyroid WBC WBC WBC WBC WBC Monitoring interval [day] 28 28 28 28 28 28 28 d) d) d) d) d) Typical amount of handled activity Tab 4. -continue 1 I Field (code) 5 5 5A 6 6 6 6 6 6 6 6 6E Nuclide "Zr uomAg MZn '"Sm 1J1I luIn •Allnpi "Ga '"'Am 1J/Cs "5I depl. U «) Compound (found or assumed) EDTMP hippurate, chloride chloride citrate oxide nitrate chloride Bengal Rose chloride Assumed Pathway (ing./inhal.) inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. AMAD [um] assumed 1 1 Monitoring method WBC WBC WBC WBC WBC WBC WBC WBC WBC WBC Thyroid Air'* Monitoring interval [day] d) d) 365 183 183 183 183 183 91 -183 185 91 7 Typical amount of handled max 30MBq max 3-6 150 400 lGBq 370 max activity 50GBq 19GBq GBq GBq GBq MBq 37 GBq Tab 4. -continue 2 Field (code) 7 7 7 7 7 7 7 7 7 7A Nuclide "Ga 155I 131I ^"Tc ^Sr «y 153Sm 186Re JOl-pi radon progeny Compound (found or assumed) Assumed Pathway (ing./inhal.) inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhai. inhal. inhal. AMAD [um] assumed Monitoring method Thyroid Air Monitoring interval [day] 14-19 Typical amount of handled activity 82MBq 30MBq 40 GBq 70GBq 150MBq 2.2 GBq 2 -7GBq 15 GBq 85MBq Tab4. - continue 3 Field (code) 8 8 8 8 8 8 9 9 OO Nuclide MZn '"In *"T1 *"T1 *"Pb compound compound HTO Compound (found or assumed) Assumed Pathway (ing./inhal.) inhal. inhal. inhal. inhal. inhal. inhal. inhal. inhal. AM AD [|im] assumed Monitoring method WBC WBC WBC WBC WBC WBC WBC Air Monitoring interval [day] 365 365 365 365 365 365 365 ») Typical amount of handled activity d) ... Special monitoring - after return from NPP (Czech Republic, other country) e) ... Found during previous time only f) ... During operation only i f 21st RHD Jasná podChopkom SK98K0360 139 CASES OF OLD INTERNAL CONTAMINATION WITH M1AM. Irena Malátová, Štěpánka Foltánová, Věra Bečková National Radiation Protection Institute, Šrobárova 48,100 00 Praha 10, Czech Republic Abstract. Altogether 14 persons with previous record of internal contamination with 241 Am were invited to participate in the study of kinetics of 241Am in the human body. One person has been participating since 1995 already, another 6 persons replied and agreed to co-operate with us. Whole body counting, skeleton counting and bioassay were performed. Special attention is paid to the calibration of whole body counter for the skeleton counting. Different approaches, using skull and knee phantoms and also Monte Carlo calculations were used for this purpose. The method of the determination of 24IAm activity in biological samples by radiochemical separation and by alpha spectrometry underwent international intercomparisons. The work is supported by the Grant NJ 4965-3 of the Internal Grant Agency of the Ministry of Health of the Czech Republic. 140 SK98K0361 21aRHDJasná podChopkom TISSUE FREE WATER TRITIUM SEPARATION FROM FOODSTUFFS BY AZEOTROPIC DISTILLATION Florentina Constantin, Ariadna Ciubotaru, DecebalPopa Inspectorate of Public Health of Bucharest, Romania 1. INTRODUCTION The actual tritium content in the environment is due to both natural production by cosmic ray interactions in the upper atmosphere and to varying amount of man- made tritium. Tritium, the radioactive isotope of hydrogen, is one of the most important radionuclide released to the environment during the normal operation of a typical nuclear power reactor. Tritiated water constitutes almost the entire radioactivity in liquid waste discharged to the environment from heavy water reactors. Therefore, monitoring tritium in air, precipitation, drinking water and vegetation is necessary for assessing the environmental impact of tritium releases in gaseous and liquid effluents from nuclear plants. The use for agricultural purposes of river water receiving releases of tritium, results in contamination of irrigated crops and then in the contamination of the food chain. ? Tritium is everywhere in living matter, has a long physical half-time (T= 12.26 y) and is known to follow protium pathways in biological material. Tritium in liquid or solid biota samples can not be easily determined quantitatively by any kind of nondestructive analysis because its beta radiation energy (E =18 keV) is so weak that the most part is absorbed in the sample. Therefore, most non-aqueous samples must be treated chemically or physically to be suitable for counting. The choice of preparation method will depend on the type of samples, number and quantities of samples to be processed and availability of measurement equipment. 2. ANALYTICAL PROCEDURE a) Sample treatment The analytical procedure of measuring tritium in biota samples requires two steps of separation: first, tissues water separation by azeotropic distillation and subsequently, oxidative tritium separation from remaining organic tissue, usually by combustion. The most part of tritium in plants is concentrated in tissue water. Therefore, all foodstuffs consumed are assumed to be about 85-90 % water and also, it can be assumed that the organically bound tritium concentration represents only 10-15 % from the total activity of tritium. Azeotropic distillation is adequate for separating water from samples with high content of water (80-90%) as vegetables, fruits, food grain plants (maze, wheat, rice), soil. st 2I RHD Jasná pod Chopkom 141 The biota samples, especially those with low tritium concentrations compared to the tritium concentration in air moisture, should be stored in sealed packages for minimizing the exposure of the sample to air. When the time between collection and analysis is over few hours, the samples must be frozen to preserve them until analysis and to prevent condensation of air moisture on the samples or looses by evaporation. The separated water will be stored in well-sealed glass containers for avoiding the exchange between the water and tritiated atmospheric moisture from laboratory. b) Principle of method Azeotropic distillation is an analytical technique, which is based on distillation of a solvent with water containing tritium. The solvent forms an azeotropic mixture with water that has a constant boiling point and whose vapor has the same composition as the liquid. The aromatic hydrocarbons as benzene, toluene and xylene are extensively used as solvents because: their solubility in water is so small (0.06%) than no significant error is introduced in gravimetric or volumetric measurement of separated water; are produced from petroleum and thus contain no tritium; hydrogen exchange between water and these solvents may be ignored. Properties of benzene and toluene for azeotropic distillation of water: Boiling Azeotrope Hydrocarbon Point Boiling Relative amount of water % °C Point °C AtBP At 20 °C Upper Lower Layer Layer Benzene 80.07 69.4 8.9 8.0 0.06 99.94 Toluene 110.56 85.0 20.2 18.0 0.05 99.95 The lower boiling point of benzene is an advantage for biota samples, which may be heat sensitive. However, processing of samples using benzene will require a long time because the concentration of water in the vapor phase is 8.9% as compared to 20.2% in toluene. Consequently, the tissue water from food samples analyzed in this experiment was separated by azeotropic distillation in toluene. b) Work procedure The food sample sliced in thin and small pieces and the solvent are placed in the distillation flask and refluxed until all the water is distillated. The distillate containing 99.93% water in the lower phase is collected and the phases are allowed to separate. To avoid the exchange of water between sample and the moisture of the air during the distillation, the air inlet of the distillation apparatus is connected to a drying tube. 142 21" RHD Jasná pod Chopkom The distillation is complete when the temperature arise about 110 °C. Normally one hour is required to extract 20 ml of water from most media. The first 10 ml of azeotrope collected should be discarded for removing the light organic from the system. Quantities of food samples and toluene required obtaining 15-20 ml of water: Type of Humidity Sample weight Toluene Water sample % g ml extracted ml Tomatoes 93-96 50 70 20 Potatoes 75-80 70 90 20 Cabbage 88-92 50 70 20 Cucumbers 90-96 70 70 20 Onion 80-87 60 80 15 Carrots 85-89 50 70 15 Apples 80-85 100 100 20 Soil 80 150 250 15 3. MEASUREMENT TECHNIQUE In generally, tritium in environmental samples, especially in foodstuffs, has a low-level activity, which requires an adequate low level measurement. Tritium in tissue water is assayed by liquid scintillation counting using well- established quenched correction methods. Various factors may affect the low level counting of tritium by liquid scintillation as: counting vials, nature and geometry of the vial, the caps, and the scintillation cocktail. Chemiluminescence and phosphorescence influence the counting rate of background and can be avoid by cooling the samples and keeping them under total darkness. The samples are prepared by transferring a known volume of separated water to a counting vial containing a scintillation cocktail and then are counted using a Liquid Scintillation Spectrometer. Into a low potassium glass counting vial, a mixture of 8 ml of separated water and 12 ml of scintillator cocktail is introduced. Blank samples consisting of dead water (less than 1.2 Bq/1) are prepared in the same way as the samples. As dead water is used deep water exposed to a natural radiation only. When the supply of tritium free water is limited, a substitute was considered. The standard scintillation solution was quenched with CCl». Varying the amount of CCl» it was found that the simulated background has the same counting efficiency and similar spectrum as the water containing. The standard counting time for all samples is 600 minutes. Calibration Standards are prepared from dilutions of Certified Trhiated Water Standards. 2ľ RHD Jasná pod Chopkom 143 4) CONCLUSIONS The mean value of the tritium concentration in tissue water from foodstuffs is about 6-12 Bq/1 very similar to the tritium mean concentration measured in the surface waters of the area where the samples have been collected (about 12 Bq/1). Therefore, the tritium content in the water fraction of the food samples can be considered in equilibrium with the local environmental water sources. The azeotropic distillation it is an accessible separation method which does not need a sophisticated and expansive distillation apparatus. It is a fast method of separation tissue free water from foodstuffs being very important in the surveillance activity of the environmental within a Nuclear Electric Plant. It is suitable for processing a small quantity of samples and for a production type facility when a large number of samples must be processed because the solvent can be purified and reused. The azeotropic distillation has some limits being used to separate water from samples with high content of water (85-90%) and a simple chemical structure as: vegetables, fruits, cereal, soil, vegetation. According to the results obtained, the organic constituents of milk, wine, meat (casein, lactose, milk fat, alcohol, esters) may enhance the chemosorption of tritium on through exchange organic hydrogen as -OH, -SH, -NH, -COOH with tritium. Also, the tissue water separation by azeotropic distillation is not complete and can not guarantee the absence of the vaporization isotope effect of the HTO/H O system. However, the azeotropic distillation is the preferred method of the water extraction from food samples, which makes it useful for study the tritium transfer from soil to foodstuffs. 144 SK98K0362 2ľRHD Jasná pod Chopkom THE AIR CONTAMINATION BY 137Cs AND 7Be IN THE TERRITORY OF SLOVAK REPUBLIC H. Cabáneková,M. Vladár Institute of Preventive and Clinical Medicine, Limbová 14, 833 01 Bratislava, Slovak Republic Introduction The Radiation Monitoring Network of Slovak Republic (RMN) was established in the year 1993 for control of the radiation situation in the territory of Slovakia. The RMN works in two regimes: first as routine operation aimed for early detection of possible accident and second as intensive operation aimed for evaluating the consequences of such an accident. This report presents the results from measurements of aerosol samples collected on the Measuring Points of Air Contamination (MPAC). The activity concentrations of 137Cs and 7Be in air during the years 1993-1996 and the annual committed effective doses due to inhalation are summarized. Beryllium-7 is a relatively short-lived (Jm = 53.3 days) naturally occurring radionuclide of cosmogenic origin, which is formed by spallation processes of light atmospheric nuclei such as carbon (Z=6), nitrogen (Z=7) and oxygen (Y=8) after absorbing protons or neutrons, from the primary componente of cosmic rays [1]. Cesium-137 is a long-lived (Jm = 30.14 years) artificial radionuclide coming from higher levels of atmosphere and from resuspension of the original fallout on the ground surface. 137 Cs in the stratosphere and troposphere is originated from the nuclear weapons tests and from the nuclear accidents. Materials and methods Regular collection of aerosol samples for routine environmental air monitoring has been provided in six Measuring Points of Air Contamination from March 1993 through December 1996. MPAC were located on the six stations of the Hydrometeorological service and all had identical parameters for installation high volume air samplers and same conditions for collecting aerosols. The concentrations of Beryllium -7 in air was estimated only for one Measuring Point at Hurbanovo ( West Slovakia, 47°52' N, 18°12' E). The high volume air samplers, typeVAJ 01, air flow of 200 m3.h4 were set lm above the ground in an appropriate distance from any impediment (houses, trees, etc.). For aerosol sampling filters type FLPES PC-9A PND 5913388 have been used. The active surface of the filter was 0.35 m2 and the collection efficiency for particles up to 0.2 fxm was greater than 99%. The sampling of atmospheric air was provided usually at the beginning of each month from March 1993 through December 1996, and the length of the collection period was approximately 7 days. During that time, using the sampler VAJ 01, the total volume of air samples was about 36 000 m3. The exposed filters were measured in the form of pellets created by pressing under the pressure of 400 kp.cm'2. The diameter of pellets was 6.4 cm and their height was 1.6-1.7 cm, depending on the quantity of the dust particles absorbed on the filter. 21" RHD Jasná pod Chopkom 145 The aerosol filters, after each sampling, were measured by gamma-ray spectrometry, using a HPGe detector PGT (high resolution 1.9 keV at 1.33 MeV for 60Co, relative efficiency 33 %) and a multichannel analyzer. The measuring time was usually 150 000- 200 000 s. The count rates in the photopeak of 7Be (447 keV) and 137Cs (661 keV) were corrected for the background of the measuring system and for the effect of self-absorption. The calculated activity concentrations of 7Be were corrected for decay reason. The uncertainty of the estimation of count - rate in the photopeak was < 1.5% for 7Be and < 25% for 137Cs. For the "pellets geometry" of measurement, the standard deviation of the efficiency was 2.5%. Committed effective dose due to inhalation of 137Cs and 7Be in air during one year was calculated as follows: Eh where Eh = committed effective dose, [SvJ l37 7 3 av = average activity concentrations of Cs or Be in air during one year, [Bq.m ] B — breathing rate, [m3.a~'] DCF — dose conversion factor [Sv.Bq1] Results and discussion Concentrations of 137Cs and 7Be in surface air from March 1993 through December 1996 in MPAC Hurbanovo are shown in Fig. 1. and the concentration of 137Cs from all MPCA are shown in Fig. 2. Cs-137 n Be-7 Modng average (2 months period) 1.0BO1 1.0BOO 12 24 36 Months after January 1,1993 Fig. 1. Cesium-137 and Beryllium-7 concentrations in ground level air from March 1993 to December 1996; (MPAC Hurbanovo, West Slovakia). 146 2ľ RHD Jasná pod Chopkom Cs-137 D Be-7 -Moving average (2 months period) Exponential trendline .= 1.0BCO —^r-- 1.OE-O1 — 1.0E-Q2 o 12 24 36 Months alter January 1,1993 Fig. 2. Cesium-137 and Beryllium-7 concentrations in ground level air from March 1993 to December 1996; (137Cs: average value at Slovak Republic, 7Be: MPAC Hurbanovo, West Slovakia). During that period the concentrations of 137Cs in air varied between 0.4 and 13.8 uBq.m"3 [2] (Tab. 1.), that is in the average seven orders of magnitude lower than the maximum value ~ 6.0 Bq.m"3 measured on 1 May 1986 in the territory of the Slovak Republic as reported by UNSCEAR (1988) [3]. Table 1. Activity concentration of Cs-137 in air during 1993 -1996 Year 1993 1994 1995 1996 Nr. of samples 23 29 29 32 Nr. of samples >MDA 7 16 20 17 Minimum 3.4 1.1 0.5 0.4 Maximum 9.1 13.8 7.3 2.9 Average 5.8 4.5 1.7 1.4 The average concentrations of 137Cs in air show a significant decrease during March 1993 through December 1996 (Fig. 2.). That decrease reflects a removal half- time of about 15 months. This removal half-time is in good accordance whit the data reported by Papastefánou et.al. (1996) [4]. The annual committed effective doses due to inhalation of l37Cs in air, according to the age [5], varied between 0.02 and 0.22 nSv from March 1993 through December 1996(Tab. 2.). 21st RHD Jasná pod Chopkom 147 Table 2. Calculation of annual committed effective doses due to inhalation of Cs-137 Parameter Ago Units 1993 1994 1995 1996 Activity concentration of Cs-137 in 3 5.8E-O6 4.6E-06 1.7E-06 1.4E-06 air IBq.m- ) <1 !m3.a-1] 1388 1388 1388 1388 Breathing rate <10 5361 5361 5361 5361 10+ [m».a-1] 8515 8515 8515 8515 <1 8.1E-03 6.2E-03 2.4E-03 2.0E-03 Intake of Cs-137 <10 [Bqa-1] 3.1E-02 2.4E-02 9.2E-03 7.6E-03 10+ [Bq.ff1] 5.0E-02 3.8E-02 1.5E-02 1.2E-02 <1 [Bq.SV1] 8.8E-09 8.8E-09 8.8E-09 8.6E-09 DCF" <10 [Bq.SV1] 3.7E-09 3.7E-09 3.7E-09 3.7E-09 10+ [Bq.SV1] 4.4E-09 4.4E-09 4.4E-09 4.4E-09 <1 [nSv.a"1] 0.07 0.05 0.02 0.02 Comm. effective dose <10 [nSv.a"1] 0.12 0.09 0.03 0.03 10+ [nSv.a-1] 0.22 0.17 0.06 0.05 •>DCF = Dose conversion factor for inhalation [5] 7Be concentrations in surfece air from March 1993 through December 1996 are presented in Fig.l and Fig. 2. During that period the concentrations of 7Be in air varied between 1.0 and 7.7 mBq.m'3 with an annual average of 2.8 mBq.m'3, which was unchanged. This average value is consisted with the value of 3.0 mBq.m'3 reported by UNSCEAR (1982) [6]. Highest concentrations of 7Be in air were usually noted during the summer period. The annual committed effective doses due to inhalation of 7Be in air varied, according to age [5], from 1.1 to 1.8 nSv (Tab. 3.). This values were significantly lower than the annual committed effective doses due to ingestion of 7Be in food and water as was reported by UNSCEAR (1993) [7]. Table 3. Calculation of annual committed effective doses due to Inhalation of Be-7 Parameter Age Units Value 3 Activity concentration of Be-7 in air IBq.m- ] 2.8E-03 <1 [m3.*1] 1388 Breathing rate <10 [m3.a-1] 5361 10+ [m3a-1] 8515 <1 [Bq.tf1] 3.8E+00 Intake of Be-7 <10 [Bq.a1] 1.5E+01 10+ [Bq.a-1] 2.3E+01 <1 [Bq.SV1J 2.8E-10 DCF* <10 [Bq.SV1] 1.2E-10 10+ [Bq.Sv-1] 5.0E-11 <1 [nSv.a"1] 1.1 Comm. effective dose <10 [nSv.a1] 1.8 10+ [nSv.a1] 1.2 *DCF = Dose conversion factor for inhalation [5] 148 21st RHD Jasná podChopkom References [1] PAPASTEFANOU, C, IOANNIDOU, A.: Aerodynamic size associations of 7Be in ambient aerosols, J. Environ. Radioactivity, 26,1995,273-282 [2] CABÁNEKOVÁ, H., VLADÁR, M., DOBIÁŠOVÁ, N., ĎUREC, F.: Volume activities of 137Cs in air during years 1993-96 on the territory of Slovakia, Proceedings the IRPA Regional Symposium on Radiation Protection, Prague 1997, 358-359 [3] UNSCEAR 1988. Sources, Effects and Risks of Ionizing Radiation, Scientific Committee on the Effects of Atomic Radiation, UN, New York [4] PAPASTEFANOU, C, MANOLOPOULOU, M., STOULOS, S., IOANNIDOU, A.: Behavior of 137Cs in the Environment one Decade after Chernobyl, Radioecology 4, 1996(1), 9-14 [5] SAFETY SERIES No. 115, International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, International Atomic Energy Agency, Vienna, 1996 [6] UNSCEAR 1982. Ionizing Radiation: Sources and Biological Effects, United Nations Scientific Committee on the Effects of Atomic Radiation, UN, New York [7] UNSCEAR 1993. Sources and Effects Ionizing Radiation, United Nations Scientific Committee on the Effects of Atomic Radiation, UN, New York 21 RHD Jasná pod Chopkom o l^gg K0363 7Be IN AMBIENT ATMOSPHERE AND EV INDOOR AIR František Ďurec, Alžbeta Ďurecová, Ľudmila Auxtová, Eduard Gombala State Institute of Public Health, Cesta k nemocnici 1, 975 56 Banská Bystrica, Slovak Republic INTRODUCTION 7 Be, relatively short lived (t/2 = 53.3 d) natural radionuclide that is produced in the upper troposphere and in stratosphere has been widely used in studying of aerosol transport in the outdoor atmosphere[l-3]. After formation by spallation reactions with nitrogen and oxygen, 7Be quickly attaches to aerosol particles and reaches the surface air via transport processes in the atmosphere. 7Be as well as other naturally occurring radionuclides participate in formation of aerosols with 0.29 to 1.18 \xm diameter, which is major reservoir of pollutants in the atmosphere [4]. From this point of view 7Be is useful tool in studying of aerosol transport from ambient atmosphere into the indoor environment. This short paper summarizes part of the results of the measurements of 7Be concentrations in outdoor and indoor air. EXPERIMENTAL The sampling site is located in Banská Bystrica. Aerosol particles are collected using two air sampling pumps (flow rate about 401/min) on cellulose filters (PC-S, SLZ Hnúšťa) with 4.0 cm diameter. The outdoor air inlet is situated at height of 6 m above ground, behind the window. The indoor air inlet is situated at a height of 1.8 m in center of room. Collection period for one filter is 7 day. After drying and weighing the aerosol mass is evaluated. Two fitters are measured for 7Be together as one sample and the total sampling period is than 2 weeks. 7Be activity is measured with HPGe detector (p-type, relative efficiency 24 %). The counting time is 3600 min. One outdoor and one indoor activity size distributions were measured using high volume cascade impactor (Sierra Andersen, Model 236, 6 stages, flow rate of 0.565 m3/min) too. RESULTS AND DISCUSSIONS In Fig. 1 and Fig. 2 the time series of outdoor and indoor 7Be air concentrations and their ratio are shown from January 1997 to the end of August 1998. Average outdoor and indoor 7Be concentrations are 2.5 mBq/m3 and 1.7 mBq/m3 respectively. The time series of aerosol masses and their ratio are presented in Fig. 3 and Fig. 4. As can be seen in the figures, the ratio of 7Be concentrations is higher in winter period. This could be explained by fact that air exchange between outdoor and indoor air is lower in winter. Differences between 7Be ratio and mass ratio can be explained by fact that 7Be is attached to small particle and with assumption of faster settlement for coarse aerosol particles in indoor air. This assumption is in good agreement with results of measurements by cascade impactor. The 7Be activity size distributions of aerosols for 150 2 ŕ RHD Jasná pod Chopkom outdoor and indoor air are shown in Fig. 5. In Fig. 6 are shown the specific activity of 7Be distributions. The AMAD (activity median of aerodynamic diameter) of 7Be aerosols for outdoor air is 0.95 urn. In Fig 5 can be seen the shift of the size distribution of indoor 7Be aerosols to the smaller diameters. CONCLUSIONS Continuous aerosol sampling of outdoor and indoor aerosols over a long period showed that the 7Be activity ratio is higher in winter period than in summer period. The measurements by cascade impactor showed the shift of 7Be size distribution in indoor air to submicron-size region. Our future work is aimed to explanation of this fact. ACKNOLEGMENTS The authors would like to thank Mr. Rulík, NRPI Prague for help in measurements by cascade impactor. REFERENCES 1. Yoichi Ishikawa, Hiroshi Murakami, Tsutomu Sekine, Kenji Yoshihara: Precipitation Scavenging Studies of Radionuclides in Air Using Cosmogenic 7Be, J. Environ. Radioactivity, 26(1995)19 2. Dorothy M. Koch, Michael E. Mann: Spatial and Temporal Variability of 7Be Surface Concentrations, Tellus 48B(1996)387 3. Simona Talpos, Vasile Cuculeanu: A Study of the Vertical Diffusion of 7Be in the Atmosphere, J.Environ. Radioactivity, 36(1997)93 4. C. Papastefanou, A. Ioannidou: Aerodynamic Size Association of 7Be in Ambient Aerosols, J. Environ. Radioactivity, 26(1995)273 1 Vokimt»ctfvttyafB»-7ln«ir t |-#-lrefccr ) 4 JO •• ; 35 f ? : ;í ' It ». : t- • ^ Í• v. J »•' V -JČ'. >: \'••. i;' .yv * • XAST 308.97 21*97 27.1207 ZMJM Z&BJM 33OM 14J7 XL&S7 2MÍ7 Z7.1ZS7 27JM 2MS8 2U« EndofMMpIhgpwtod 2 ľ RHD Jasná pod Chopkom 151 Ffc. 3 SptcKc mm *•*«•«* In ri- ng* RHo«fspKincRin*«rMrMOl»lniirteuWoorlndMi] ŠI J • t* i : : •. i • • * •.: • * •-• •; •••• • • S : i ľ\ 1h Y~ 1.1J7 14A 3O6S7 28.9S7 2T.1U7 27^8 2&6M 2UM 1.1J7 I4J7 3QM7 S&9J7 37.12J7 27JSS 2UM 07 Q8 Ž" 1 J" 1 i 03 1 í 1 ÍQJ !•!• Q1 1 ! p 0 mím 1 152 SK98K0364 2ľRHDJasnápodChopkom RADIOCHEMICAL ANALYSIS OF ENVIRONMENTAL SAMPLES Pavel Dillinger, Margita Harangozó, Juraj Tôlgyessy1 Department of Environmental Science, Faculty of Chemical Technology, Slovak University of Technology, Radlinského 9, 812 37 Bratislava, Slovak Republic ' Department of Chemistry, Faculty of Natural Science, Matej Bel Univerzity, Tajovského 40, 974 01 Banska Bystrica, Slovak Republic The development of application of radioactive materials in research, industry, medicine, and agriculture etc., led to increasing amounts of varieties of radionuclides in the environment. This has created the demand for proper analytical methods of determining radionuclides in the environment. The major concerns of radiochemical analysis is the study of radioactivity in the environment - its nature, occurrence, behaviour, and control. Such measurement can be divided into two major types: 1. Measurements on systems of naturally occuring radionuclides where objectives include the determination of identity and amounts of natural radionuclides, and the use of this information to determine the presence and amounts of the related elements in environmental samples (e.g. U, Th, and K). 2. Measurement of man-made (artificial) radionuclides, including determination of their identity and amounts (e.g. fission products). From the practical standpoint, radiochemical analytical procedures can be divided into two groups, one that is characterized by a high sensitivity and a second that offers high speed but relatively lower sensitivity. When selecting a proper analytical method sometimes the faster method will be chosen and, other times, the more sensitive one [1-8] In Table 1 a summary of radiochemical procedures for environmental samples for various radionuclides is given. REFERENCES 1. Tolgyessy.J. andBujdoso,E.(1991) Handbook of Radioanalytical Chemistry, Vol.2, CRC Press.Boca Raton., Fl.,. 2. Tolgyessy.J. and Klehr,E.H.(1987) Nuclear Environmental Chemical Analysis, E.Horwod, Chichester. 3. Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities (1969), ANSI N 13.1, American Standards Institute, New York. 4. Tolgyessy.J. and Varga.Š. (1974) Nuclear Analytical Chemistry, Vol.3, Radiochemicaland Activation Analysis, University Park Press, Baltimore. 5. Drenham,D.H. (1982) in CRC Handbook of Environmental Radiation, Klement A.W., ed., CRC Press, Boca Raton, FL, 2. 6. Environmental Radioactivity Surveillance Guide (1972), ORP/SID-72-2, Office of Radiation Programs, U.S.Environmental Protection Agency, Washington D.C.. 7. Corley,J.P.( .Denham,D.H., Jacquish,R.E. .Michels.D.E., Olsen,A.R. and Waite,D.S.(1981) SA guide for Environmental Radiological Surveillance at U.S.Department of Energy Institutions, DOEP-0023, Office of Operational Safety, U.S.Department of Energy, Washington D.C.. 8. Harley J.H. HASL Procedures Manual, HASL-300, Health and Safety Laboratory, U.S. Atomic Energy Commission (revised annually by the Environmental Measure- ment Laboratory, U.S.Department of Energy). st 21 RHD Jasná pod Chopkom 153 TABLE 1. Summary of Radiochemical Procedures for Environmental Samples Ana- Sample Preparation, Chemical Methofof Iyte dissolution separation measurement "Na Precipitati Add Na carrier Na separated by ion p^ Coincidence on exchange,hydroxides counter scavenged, Na measured in filtrate 131j Milk Add HCHO and I I separated on anion-exchange (3-Counter carrier resin, purified by extraction; Pdl2 precipitated 137 Cs Milk,bone, Fuse-Na2CO3 or Ca precipitated;Cs separated p-Counter vegetation leach-HNO3 on NH) phosphomolybdate, ash,soil, purified by ion water exchange,precipitated as CsjPtCU 2L0pb Bone Ignite-dissolve in Pb extracted into quaternary P-Counter HBr amine;PbSC>4 precipitated and (210Bi counted) counted 2J2Th Bone Ignite at 600°C Irradiate with y-Ray neutrons.dissolve in Hcl, spectrometry of separate 223Pa by anion 2MPa exchange ^Sr Bone Ignite,dissolve with "°Y extracted into ethylhexyl HC1 phosphoric acid; impurities P-Counter removed by amine extraction wSr Food Evaporate, Sr separated, purified by P-Counter vegetation, ignite,fuse with nitrate precipitations, "V tissue Na2CO3;Soil-fiise precipitated and counted milk,soil, with Na2CO3 or water leach with NaOH,HCl; "Sr and Sr carrier added "7Cs, Sea water Acidify-HCl;added Cs removed by NH4 P-CountCT "Sr carriers phosphomolybdate, purified, precipitated as C^PtCU; Sr precipitated, ""Y daughter separated and counted Milk,water Milk-I first separated by anion Liquid exchange; milk and water-I scintillation extracted into CCI4 and counter toluene, and decolorized with UV light in presence of 2- methyl-1-butane Th-Cf Soil Treat with Actinides coprecipitated with a-Spectrometry acids.fuse- BaSO4, separated by solvent KF,Na2S2O7, extraction after adjusting dissolve-HCl oxidation states, electrodeposited for counting U Water Extraction into methyl Fluorescence of " isobutyl ketone U infused NaF Pu Soil.water, Leach with mineral Pu purified by anion a-Spectromety air.food, acids exchange; electrodeposited for vegetation counting 154 SK98K0365 21" RHD Jasná pod Chopkom THE SYSTEM FOR AUTOMATIC DOSE RATE MEASUREMENTS BY MOBILE GROUPS IN FIELD Dana Dráhová, Radim Filgas, Irena Češpírová, Marie Ejemová National Radiation Protection Institute, Šrobárova 48,100 00 Prague 10, Czech Republic Mobile groups are considered to be fastest, most flexible and most complex units suited for rapid direct measurements after a radiological accident. All possible means for the assessment of the temporal and spatial variation of the contamination should be utilised together to support the decisions best suited for the given situation. Dose rate measurement is a fest and simple method to detect radioactive contamination of the environment. The instruments are small and light, which makes it easy to take them into the areas of interest. An automatic route monitoring system can collect data on the temporal and spatial variation of the dose rate in contaminated areas. During an environmental monitoring all the measured data have to be connected with exact location. Thus, a detector, a GPS navigator (co-ordinates) and a portable computer are needed together for accurate record keeping of the data flow. Using suitable software, the dose rate data can be dispayed on digital maps and the local variability can be spotted. The advantages and disadvantages of selecting a pressurised ionization chamber (PIC) - Reuter Stokes 112, a plastic scintillator - Tesla NB 3201 and proportional counter - Eberline FHZ 62IB are compared in Table 1. The technical performance of the PIC i.e. the precision and time response of the measurements is superior. Almost equal quality can be achieved with the plastic scintillation detector. However, for field emergency measurement use of the proportional counter seems to have advantages that cannot be overlooked. Table 1 Comparison between a pressurised ionization chamber (Reuter Stokes 112), plastic scintillator (Tesla NB 3201) and proportional counter (Eberline FAG FHZ 62 IB). PIC(RSS-112) Plastic scintillator Proportional counter (NB 3201) (FAGFHZ621B) Precision 0.01 uSv/h 0.02 fiSv/h 0.03 nSv/h Time response to very fast (5s) fast (10s) slow (minutes) achieve precision Energy response > 60 keV > 30 keV > 50 keV Weight 10 kg 4kg 2 kg GPS connection External Field use good very good very good 21"RHD Jasná podChopkom 155 Based on above mentioned requirements and comparison of properties of various probes, the system for automatic dose rate measurement and integration of geographic co-ordinates in field was designed and tested in NRPI. The system consists of proportional counter FHZ 62IB by Eberline. This so-called intelligent probe can be easily connected to a personal computer. The probe measures in the energy range 30 keV -1.3 MeV with reasonable energy and angular response, it can measure the dose rate in the range 50 nSv/h - 1 Sv/h with the typical efficiency 9.5imp/s/uSv/h. The probe is fixed in the holder placed on the front mask of a car. For the simultaneous determination of geographical co-ordinates the personal GPS navigator Garmin 95 is used. Both devices are controlled by a notebook PC (486DX2 50MHZ, 340 MB HDD, 8 MB RAM) via two serial ports. The second serial port that is not quite common in notebooks can be easily realised by a PCMCIA card. The notebook is used for data transfer, processing and storing. Data stored in the notebook in the field by a mobile group can be transmitted to the assessment centre by the cellular GSM phone. In our system Nokia 2110 connected to notebook by PCMCIA card is used. The whole system is powered up from the car battery. The system is controlled by specially developed software. The software was developed in the FoxPro 2.5 environment and works under MS-DOS 6.22. It has no problems to work in Windows 95 DOS window. The software allows: • automatic determination and recording of dose rate values and synchronisation with the geographic co-ordinates in real time, • mutually independent determination of dose rate and co-ordinates which allows to measure the dose rate manually if the automatic probe fails and still have information on co-ordinates or to measure the dose rate automatically even if the GPS receiver is not available, • manual determination of dose rate and entering the data into the computer, • manual determination of co-ordinates and time and entering the data into the computer, • to define the integration time for dose rate (by time or location), • to define various probes for dose rate measurement, • to define measured quantities and relevant units and recalculation to nSv/h, • to send the data to the central database. The results of dose rate measurements obtained during route monitoring are stored in files. They can be displayed on a graphic screen, presenting the geographical distribution of the dose rate values colour coded on a map and the time sequence of the measured data. 156 SK98K0366 21" RHD Jasná pod Chopkom INTERCOMPARISON MEASUREMENT - ORAVA SOIL František Ďurec, Alžbeta Ďurecová, Ľudmila Auxtová, Eduard Gombala State Institute of Public Health, Cesta k nemocnici 1, 975 56 Banská Bystrica, Slovak Republic INTRODUCTION In 1997 our laboratory organized intercomparison measurement in gammaspectrometry. The soil samples were delivered to 20 laboratories in 4 countries. The main aim was check the laboratory practice of member of the Slovak Radiation Monitoring Network. This short summary contains the results of the intercomparison on the determination of radionuclides in soil by gamma spectrometry. DESCRIPTION OF THE MATERIAL The soil was collected in autumn 1996 near Námestovo (Orava region) from layer 0-5 cm. The air-dried sample was sieved to obtain a grain size smaller than 2 mm and manually homogenized. Aliquot parts of the material (600 g) were distributed to 21 plastic Marinelli bakers, which serve like transportation cover for material and for homogeneity testing. The homogeneity of the soil samples was tested by gamma spectrometry on each Marinelli baker. Considering the results of 40K and l37Cs measurements, the count rates did not differ significantly (significance level a = 0.05) and thus this material has been considered sufficiently homogenous for this intercomparison. RESULTS The summary of received data is presented in Table 1. The original data that were received from the participating laboratories were converted to the same units and format. Then the data were processed by the computer program STATA. Results which deviated significantly from the population were considered to be outliers and rejected if they foiled either Dixon's or Grubb's statistical tests at the significance level of a=0.05. Although data on 20 radionuclides were received, only data for 134Cs, I37Cs, 226Ra, 232Th and 40K were statistically evaluated. The results for these radionuclides are presented in Figures 1-5. Data for other radionuclides are not presented. The used half-times and yields do not differ significantly. The differences between laboratories in reported results are due to mainly by sample treatment and spectra evaluation (used efficiency). ACKNOWLEDGEMENTS We would like to thank all the participating laboratories for their co-operation in providing the data for this intercomparison. 2ŕ RHD Jasná pod Chopkom 157 TAB I. Summary of received data Miiclid : Mean " 3t.dev; Raage Outliers e , * (Mb. of considered) feq/kg] CBtyfcg] mCs 16(14) 1.05 0.36 0.31-1.85 4.19 "JCs 16(14) 59.7 1.6 57.5 - 62.7 70.5 226Ra. 16(14) 24.0 1.5 21.8-26.9 50.7 16(15) 27.3 2.8 22.4-33.1 - 16(15) 411 39 346 - 480 - Fig. 1 Cs-134 In Orava soil, 1997 Mean valu*: 1.05 Bq/kg 6 5 -- 4 P i-3 2 1 0 H 1 1 1 i 1 1 1 1 h H h 2 3 15 16 11 13 7 5 4 8 14 1 12 6 9 Lab. code Fig. 2 Cs-137 In Orava soil, 1997 Mean value: 69.7 Bq/kg 75 70 -- 65 - r> i- 60 i Í " " 55 - 50 -•• 45 -I 1 1 H H 1 1 1 h -i 1 h 2 12 16 11 14 6 3 4 1 5 7 13 9 15 8 Lab. code 158 21st RHD Jasná pod Chopkom Fig. 3 Ra-226 In Orava soli, 1997 Mean value: 24.0 Btj/Kg 60 50 - 40 -•• 30 -•• 20 - - ILulil.:.!:. 10 -I 1 1 1 1 1 1 i H H 1 1 16 4 5 3 7 13 15 11 14 8 2 12 S 1 6 Lab. code Fig. 4. Th-232 In Orava soli, 1997 Mean value: 27.3 Bq/fcg 40 3B 38 34 4 32 T j. 30 .'""T 28 : J--] 26 „•••-< , • i • -t 24 T < ' n 22 20 —i—i—i—t—i—i—i—i—i—i—i—i—i— :4U 7 15 6 11 16 12 14 13 2 3 5 1 8 9 Lab.code Fig. 5. K-40 In Orava soli, 1997 Mean value: 411 Bq/kg 520 500 480 460 i: 440 .*. 420 J? 400 m 380 360 340 320 300 280 260 H 1 1 1 1 1- 16 H 6 15 13 12 11 51 11 13 4 7 g 1 8 12 14 Lab. cade 2 ľ' RHD Jasná pod Chopkom SK98K0367 159 IN SITU GAMMA RAY COUNTING OF THE LARGE VOLUME OBJECTS USING THE LARGE AREA PLASTIC SCINTILATORS Štefan Krnáč Department of the Nuclear Physics and Technology, Slovak Technical University, Ilkovičova 3, 81219 Bratislava, Slovakia 1. Introduction Three basic ways for counting of large volume objects are at present commonly used. They are based on the detectors as follows: i) spectrometry detectors including especially semiconductor and anorganic scintilation detectors, ii) large area plastic scintilation detectors with dimension of several squared metres, 'unproportional counters andionisisation chambers. After analysis of the ways for counting of large volume objects, a best compromise seems to be using the large area plastic detectors with combination of whole spectrum processing based on the response operator method. Results of testing this method are btroduced in following sections. 2. Instruments Measurements have been performed using the two large area detectors on basis of plastic scintilator of type FHT 620C from Eberline Instr. Erlangen Germany with scintilator dimension 100 x 50 x 5 cm. The detector part has included built-in high voltage supply and photomultiplier tube. The useful energy of the detection is above 100 keV. vcm-.m&m SCRAP SOFTWARE DETECTOR #1 COUNTER DETECTOR #2 BUFFER PC«2:MCAACOJSPBC DfiTECTIONSYSTEM METRA SOFTVŕÄRB LEGEND: HVS - High Voltage Supply AMP - Amplifier PM - Photomultiplier Tube COUNTER -16 bit Counter Card -20 VS- -20V Supply ADC - AD Converter RCF - RC Filter BUFFER - 2048 Channels Memory SAMP - Spectroscopy Amplifier Fig. 1 Block scheme of the counting system Pulses from the detectors (loaded over -20 V supply level) have been fed into counters of Eberline's FHT8000 electronics and simultaneously, after separation of-20 V level by a RC filter, also into Canberra's MCA Accuspec card for followed 160 2 ľ1 RHD Jasná pod Chopkom spectrometric analysis. The obtained spectra have been analyzed by own PC software called METRA. A basic connection scheme of the counting system is outlined in Fig. 1. 2. Methods The time analysis In this connection, the Eberline's FHT8000 system for a time analysis works properly in full original extent. It allows to perform without any restriction all basic functions such as setting up alarms, background correction for searching sources under a background level, switching on the digital filter (ADF) and performance of channel logic operations. The system is first of all functioning as a very good „filter" for searching the intense gamma-ray sources situated within a large object and strongly shielded by it. However, in this configuration, only relative quantities related to a background level are to obtain. The spectrometry analysis A typical pulse height spectrum in the plastic detector of the given dimension was obtained by MCNP calculations simulating a point source of 60-Co placed perpendicularly at 100 cm from the centre of the detector and can be seen in Fig. 2. Fig. 2 Pulse height spectrum of 60-Co for plastic detector simulated by MCNP code In order to analyze the wide-spread plastic spectra, the response operator method for calculation of incident gamma-ray spectra has been employed [1]. A measurement of the spectra may be described using the matrix approximation of spectra by the following matrix equation S = KG, where S is a matrix of the physical spectra, G represents the incident spectra, and K is a matrix representing the response operator of the measuring device. The use of the method, for in situ applications, is described in [2] in detail. The operator K can be estimated by measuring point standard sources using a technique based on the scaling confirmatory factor analysis (SCFA) presented in detail in [1]. Calculation of G allows directly to evaluate the fundamental photon dosimetric quantities such as photon exposure and/or dose rates [2]. 21" RHD Jasná pod Chopkom 161 3. Calibration The spectrometric calibration of the plastic detectors has been performed using a set of standard point gamma-ray sources being placed at the centre of the detector close to its cover. Measured physical 128-channel spectra from two of these sources are plotted in Fig. 3 (137Csand ioCo). The response operator K has been constructed by the SCFA technique. Two significant factors has been included in the operator calculations as follows: i) a Compton factor, representing the fundamental photon interactions with a matter, especially Compton scattering (see Fig. 2), ii) a noise factor, originated from the fluctuations of photon energy conversion into electric pulses, especially the fluctuations of luminiscence in a scintilator. C)137-CW O 1)203040506070809013010130 0 D2)3).«506070a0S0t»tD120 Channel Channel Fig. 3 Physical plastic scintilator spectra of 137Cs and 60Co In regard to very continuous character of the regular reponse, the derivated operator being applied to the derivated physical spectra is more suitable one than the original operator. Comparison of derivatives of the measured and the two-factor modelled spectra for DET #2 is shown in Fig. 4 for 137Cs and 60Co spectra, respectively 0 €2D3040506070aogOtXI1t)120 fl2030«50e07080901D1O-B0 Channel Channel Fig. 4 Derivated measured and two-factor model spectra of 1J7Cs and *uCo A capability of the qualitative analysis by the response operator may be demonstrated via Fig. 5, where unfolding results of some physical spectra of DET#2 162 21s1 RHD Jasná pod Chopkom (see Fig. 3) are figured. It may be found that this approach allows clearly to distinguish two energy lines of 60Co. 4. Metrological verification Relevance of measuring by the system METRA has been verified by the Slovak Institute of Metrology (SIM). A set of SIM reference standards listed in Tab. 1 has been employed for comparative measurements with the metrological standard device. The comparison has been performed according to dose rate determination at 100 cm distance from the reference source. The counting time was 100 s. For the source No. 2 and 4, the direct flux of photons into the detector was collimated out of the detector. The background has been long-term monitored cca 150000 s. Whereas the pure signal over background was evaluated, a background correction is included in the error estimates which are tabled commonly with the METRA results of the comparison in Tab. 1. The energy range for dose evaluation was from 200 keV up to 1900 keV. •MO0OO 120000 , 120000 tXXJOO 80000 80000 40000 20000 0 0,1 0,4 0,6 0,9 W V» 16 19 0,1 Q4 0,6 0,9 11 14 16 19 Energy (MeV) Energy (MeV) Fig. 5 Unfolded photon fluence rate spectra of 137Cs and 60Co The obtained METRA values of metrological comparison were in good accordance with the SIM values obtained by a pressurized ionisation chamber. The dose values were calculated from the photon fluence rate spectra at 100 cm achieved by application of the SCFA response operator to the measured plastic physical spectra [2] (for instance see Fig. 5 for 14-keV-channel distribution of the photon fluence rate [pps]). Dose Statistic Systematic Total Relative Source Activity Geometry Nuclide rate error error error error No. (kBq) (nGy/h) (nGy/h) (nGy/h) (nGy/h) (%) 1 100 cm 137-Cs 413.9 29.2 ±3.11 ±1.59 ±3.49 ±11.9 2 collimated 137-Cs 413.9 1.9 ±2.78 ± 1.58 + 3.20 ±171 3 100 cm 137-Cs ca3400 327.4 + 5.46 ±1.76 + 5.74 ±1.75 4 collimated 137-Cs ca3400 19.1 ±2.92 ±1.59 ±3.33 ±17.4 5 100 cm 137-Cs 344.4 24.3 ±3.09 ± 1.6 ±3.48 ±14.3 6 100 cm 137-Cs 3.47 2.7 ±2.83 ±1.59 ±3.25 ±119 7 100 cm 137-Cs 39.98 4.5 ±2.85 ±1.59 ±3.26 ±72.9 8 100 cm 60-Co 789.1 105.5 ±5.68 ±1.63 ±5.92 ±5.6 9 - Backgrnd - 48.8 ±0.05 ±1.79 ±1.82 ±3.8 Tab. 1 Results of metrological verification (DET#2) 2 ľ RHD Jasná pod Chopkom 163 Conversion of a Ge gamma-ray to absorbed dose rate has been proposed by several investigators [2], [3], [4], [5], and [6]. As compared them, best results are achieved by SCFA method proposed in [1] and [2] and, furthermore, the method can be comfortably applied to 4096 and more channels spectra, and also to plastic scintilator spectra with very good results. 5. References [1] Š. Krnáč, P. P. Povinec, J. Radioanal. Nucl. Chem., Vol. 204, No. 1, 57-74 (1996). [2] Š. Kmáč, P. Ragan, Radiat. Protect. Dosira, Vol. 58, No. 3,217-228 (1995). [3] A.Clouvas, S. Xanthos, M. Antonopoulous-Domis, J. Silva, Health Phys. Vol. 74, No. 2,216-230 (1998). [4] H. Terrada, E. Sakai, M. Katagari, J. Nucl. Sci. and Technol., Vol. 17, 281-290 (1980). [5] K. M. Miller, L. H. Beck, Radiat. Protec. Dosim. Vol. 7,185-189 (1984). [6] G. Fehrenbacher, R. Meckbach, P. Jacob, Nucl. Inst. and Meth. A383,454-462 (1996). 164 21st RHD Jasná pod Chopkom SK98K0368 PERSONAL DOSIMETRY SERVICE IN THE SLOVAK REPUBLIC Jaroslav Compel Legal metrology services ofSR, Personal dosimery laboratory, Geologická 1, 82211 Bratislava From first January 1996 in Slovak Republic was started a new independent national personal dosimetry service on external ionizing radiation provided by the Personal Dosimetry Laboratory in the Slovak Institute of Metrology in Bratislava. At first July 1998 was this laboratory delimited from Slovak Institute of Metrology to Legal metrology services of SR. This dosimetry services is based fully on the Automated Thermoluminescent Dosimeter Card Reader System Model 6600 made in USA by Harshaw-Bicron NE, which was given to the Nuclear Regulatory Authority of SR and Slovak Institute of Metrology in Bratislava on an International Technical co-operation project piloted and financed by International Atomic Energy Agency, In this time we have two complete TLD Readers Harshaw 6600 and we calibrate and evaluate TLDs of three different type of all body personal dosimeters: 1) beta-gamma TLDs type 0110 with 2 element TL chip type 100 (LiF:Mg,Ti 3,2x3,2x0,3 8mm) fixed in two teflon windows in a aluminium card in a plastic holder with 1000 mg/cm2 PTFE hemisphere thickens (olOmm) to measure the individual dose equivalent penetrating at a the depth 10 mm (Hp(10)) and individual dose equivalent superficial Hs(0,07) at recommended depth of 0.07 mm (type of this holder is 8814). 2) beta-gamma TLDs type 1111 in holders type 8805 with 4 elements of TL chips type 100 to measure both Hp(10) and Hs(0.07) plus lens of eye dose Hle(3). 3) For measurement of doses in mixed neutron-beta-gamma fields we have more neutron beta gamma dosimeters type 7776 in holders type 8805 with 4 elements of TL chips: three TLD 700 and one TLD 600 LiF chip. Except of them we have two type of additive extremity dosimeters: 1) finger dosimeters and 2) wrist strap (bracelets) dosimeters to measure direct skin doses on the fingers or on the hands of the workers in radiation fields. The periods of the dosimeters evaluation are 3 months or 1 month for the basic (the all body) beta-gamma dosimeter with 2 element of TL-100 material and 1 month for all other type of dosimeters. Now we monitors about 6000 persons by all body TL dosimeters and about 720 persons by additive finger of wrist strap dosimeters. These are about 50000 measurements in a year. Each dosimeter has a oven identifications number and bar code. To each monitored persons is reserved a pair of dosimeters which are changing one to the another once more per one or three month expositions period. Each dosimeter we calibrate once more per two years, the reader once more per year. The TLD system is verified once more per 2 years by one series of TLDs irradiated at the different dose by different energy. For each type of dosimeter we make and use different calibration. 21" RHD Jasná pod Chopkom 165 We use different calibration factors (RCF) for the measurements of TLDs instead of the real exposition time of the dosimeters, to simply eliminate the time fading factor. The reader we calibrate by the calibration cards which presents the medium „gold dosimeters" from all TLD series of one type of dosimeters. These we have irradiated in holders by Cs-137 4 or 8 pc at once on the 30x30x15 cm PMMA phantom (with a 5 mm Plexiglas shield to eliminate the secondary electrons). The dosimeters for practical using we calibrate after irradiation in the holders on the same PMMA phantom (12 pcs at once) with Co-60 source moderate with two 2 cm thick Pb shield in 5000 mm distance from source. We1 don't use Plexiglas shield to eliminate the secondary electrons if we make relative measurements to generate element correction factor for each elements of each dosimeters and to measuring we use a special calibration of reader by calibration cards irradiated by the same source (Co-60) and in the same geometry without plexiglas shield. As so as we calibrate the reader for measuring the finger and bracelet dosimeters: calibration dosimeters (4 or 8 at once) are irradiated by Cs-137 on the finger PMMA phantom ( TLD100 - Background J.Compel, LOD SLM SR TimeFading TLD100 1000 J.Compel.LOD SLM SR 21"RHD Jasná podChopkom SK98K0369 167 METHODS OF ASSESSMENT OF WHOLE BODY 241AM CONTENT. Štěpánka Folíánová, Irena Malátová, Jan Klisák National Radiation Protection Institute, Šrobárova 48,100 00 Praha 10 Introduction In vivo spectrometry remains a major tool for individual monitoring to assess internal contamination. When low energy emitter is subject of interest, measurements of organs or individual tissues are usually used rather than whole body counting. For detection of bone-seeking radionuclides, measurements over the regions like head and knees are suitable. This paper discuss an influence of different skull phantoms on efficiency of the measurement. Description of some methods of an assessment of the 241Am content in the human skeleton from measurements performed over long bones of the human body is also offered. Description of calibrations with different skull phantoms There were four different skull phantoms used for calibration. The measurements were carried out in geometry with two LEGe detectors of 20 cm square active area each, situated over temporal regions of the skull (one detector on each side). The phantoms used for calibration were: The head phantom of the US Transuranium and Uranium Registry (USTUR) that is a part of four pieces bone phantom (head, thorax, arm and leg) and contains a half skeleton from a voluntary donor, who incurred an acute accidental exposure to 241Am via a wound. The other half of the skeleton was radiochernically analysed for 241Am. This analysis, performed 25 years after intake, showed a ratio of 241Am concentrations in compact/cancellous bone 0,61±0,14 that indicates an approach to uniform americium distribution in bone. Bones of this phantom were covered with appropriate amounts of tissue-equivalent material to provide attenuation of photons similar to that of the soft tissues that cover bones in the human body. The skull phantom of University of Cincinnati - UCSKULL94 that was fabricated using a real human skull with known amounts of 241Am activity placed on all external and internal surfaces of the skull. Physiognomy of tissue equivalent facial structures were approximated by covering the skull with tissue equivalent material of thickness approximating that for the human face. Other organs of the head, such as brain, tongue, eyes etc. were constructed from tissue equivalent material. The skull phantom of Bundesamt fiir Strahlenschutz - BfS that used a real human skull with 241Am artificially loaded on its inner and outer bone surfaces and covered with tissue-equivalent wax. Brain, tongue, eyes etc. were reconstructed from tissue equivalent material. The inside of the skull was filled with tissue equivalent material in the form of small spheroids. The National Radiation Protection Institute skull phantom - NRPI phantom that also used a real human skull on which inner and outer surfaces, except the facial region 168 2ľ'RHD Jasná podChopkom of the skull, a planar source of defined 241Am activity was shifted to simulate homogeneous distribution of 241 Am activity on external and internal surfaces. Unlike other phantoms NRPI phantom had no mandible. The phantoms used for calibration and efficiencies measured are presented in tab.l. This table reveals reasonable agreement in efficiencies measured, when using UCSKULL94 and USTUR skull phantoms and a discrepancy of about 50 per cent, when compare NRPI and BfS to UCSKULL94 and USTUR skull phantom efficiencies. This can be particularly caused by the size of the skulls. The NRPI and BfS skull phantoms looked visually smaller than "American" phantoms, that can be caused by different sex or human race of the subjects, who devoted their skulls for research purposes. The efficiency measured with NRPI phantom must be further corrected for mandible missing (about 10 per cent less), for the soft tissues overlying the skull (about 9 per cent less) and also for 241Am activity that is deposited in facial regions of the NRPI skull phantom (about 10 per cent less) [2]. If all these corrections are considered, final efficiency will be reduced to about 75 per cent. Results of calibrations mentioned in previous two paragraphs are in a good agreement with results of measurements carried out with the same skull phantoms in other laboratories in Western Europe, using also a different calibration geometry. The difference is now further investigated in much more details [1]. Compare the efficiency per square centimetre of the detector window area using the UCSKULL94 phantom in geometry with two LEGe detectors over temporals it is approximately a half using the same skull phantom and two phoswich detectors of 15 cm diameter in the same geometry [3]. Description of calibrations with different leg and arm phantoms For the assessment of amount of 241Am deposited in long bones of the human body (tibiae and fibulae, ulnae and radii) there were several different methods used. The method using the USTUR BPAM-001 leg and arm phantoms for calibration is one of the most proper, mainly because of nature of phantoms. As was said above a unique of these phantoms is in deposition of 24lAm in bone matrix, which had accumulated into the bone via natural metabolic processes after accidental intake. Also anatomical form of tissue equivalent material that substitutes soft tissues gives similar photon attenuation. Calibrations with these phantoms were performed with two (HPGe and LEGe) detectors in three (two) positions at 15 cm intervals starting 7,5cm under knee (elbow). Windows of detectors were positioned rectangularly to the longitudinal axis of the calf (forearm) 10,5 cm from this axis. Other method to asses 241Am activity deposited in long bones of the calf and forearm is based on idea to make measuring geometry as simple as possible. In this way, the long bones were simulated first with a linear source of 241Am, further approximated by moving a point source of 24IAm in 5 cm intervals along the line, situated in the longitudinal axis of the calf and forearm. There was expected, that this simplification will provide lower estimated 241 Am activity deposited in the skeleton of a real case of internal contamination with 24IAm, than it is a real skeletal burden. This would be because of photon attenuation in bone volume. Also attenuation in soft tissues overlying bones is not taken into account. Long bones of the calf and forearm were also simulated by moving a plaster block (15x2,8x3,8 cm) of homogeneously distributed 21" RHD Jasná pod Chopkom 169 241Am activity in its volume, situated in the longitudinal axis of the calf and forearm. In this case, there was expected to receive a higher estimate of 241Am skeletal deposition, than a real skeletal burdea This would be because of higher photon attenuation in plaster block, than in a real bone and overlying soft tissue together. Aim of this idea was to create an interval of estimates in which a real skeletal burden of a person internally contaminated with 241Am would lie [4]. Calibrations with both linear and block sources of 241Am were performed in the same detector geometry as described for the arm and leg USTUR BPAM-001 phantoms in previous paragraphs. As it can be seen from table 2, the efficiencies measured, when using the arm and leg USTUR BPAM-001 phantoms are quite within the interval of efficiencies using linear and block sources of Am. Final estimates of activity deposited in the skeleton of a person internally contaminated with 241Am can be calculated, when assume, that long bones of the calf and forearm contain 5,74 and 1,52 per cent of 241Am activity, that is assumed to be homogeneously distributed in the whole skeleton [5]. The method described above using linear and block sources of 241Am has no demand to be accurate in assessing 24IAm activity deposited in the long bones of the human body. It only showed an easy way to receive estimates of 24lAm skeletal burden, that are comparable to estimates, when using much more sophisticated phantoms. Table 1. Calibration factors using different skull phantoms in geometry with two LEGe detectors positioned over temporal regions of the skull (one detector on each side) Skull phantom Efficiency USTUR 0.0055 UCSKULL94 0.0068 NRPI 0.0133 (0,0100)* BfS 0.0122 *reduced efficiency Table 2. Calibration factors over the calf and forearm using linear and block sources of 241Am and arm and leg USTUR BPAM-001 phantoms Phantom (source of Calibration factors from Calibration factors from M1Am) used for measurements over measurements over calibration the calf the forearm Linear source of Z41Am 0.0284 0.0113 Block source of 241Am 0.0183 0.0083 USTUR phantom 0.0213 0.0119 References: 1. Rtthm, W., Kônig, K., Malátová, L, Doerfel, H., Foltánová, Š. et. al.: Intercomparison Exercise for the Determination of241 Am in the Human Skeleton, Workshop Intakes of Radionuclides, Occupational and Public Exposure, 15-18 September 1997, Avignon, France. 170 2 ľ' RHD Jasná pod Chopkom 2. The U.S. Transuranium Registry Report on the 241Am Content of a Whole Body, Special Issue. Health Phys. 49,1985. 3. Hickman, D. P., Cohen, N.: Reconstruction of a Human Skull Calibration Phantom Using Bone Sections from an 241Am Exposure Case. Health Phys. 55, 59-65, 1985. 4. Klisák, J.: Návrh metodiky pro studium distribuce osteotropních radionuklidů v těle člověka. Diplomová práce KDAIZ FJFI ČVUT, 1998. 5. Kathren, R. L.: The United States Transuranium and Uranium Registries: Overview and Recent Progress. Radiat. Prot. Dosim. Vol. 26 No. 1/4 pp. 323-330,1989. The work was partially supported by the Grant NJ/4965-3 of the Internal Grant Agency of the Ministry of Health of the Czech Republic. Ill 2ŕRHDJasnápodChopkom SK98K0370 171 CONTAMINATION OF A NEUTRON GENERATOR FACILITY BY TRITIUM II. M. Tomášek Institute for Nuclear Physics of the Academy of Sciences of the Czech Republic Department of Radiation Dosimetry Na Truhlářce 64, 180 86 Praha 8 INTRODUCTION In the building of the Department of Radiation Dosimetry of the INP CAS a neutron generator (Irelec, France) had been installed until the year 1993 when it was moved to the Czech Institute of Metrology. The generator works on the principle of reaction of accelerated deuterons with tritium target. Handling these targets of an activity of the TBq order which consisted of a metal plate on which tritium is fixed usually in titanium film can cause contamination of the workplace by this radionuclide. Particularly used targets are a source of contamination as they are exposed to sufficient heat and mechanical stress during operation of the generator [1]. Tritium released to the atmosphere is present mainly bound with tritium particles which can contaminate atmosphere of the workplace as atmospheric aerosol or in the gaseous form as HT or HTO vapours [2]. Whilst in the previous communication [3] we examine contamination of the workplace during operation of the neutron generator, the present paper is aimed at studying the decrease in workplace contamination after removal of the generator. The reason of paying such a great attention to tritium is because its use is supposed in fusion reactors in which it will represent the most significant radiological risk. PROCEDURES Tritium volume activity in humidity was measured in order to determine changes of contamination of the workplace. Two types of samples were prepared: discrete ones for evaluation of chosen areas of the building and cumulative ones which were taken continuously from the surrounding atmosphere over the roof of the Institute in the connection with the programme of monitoring of 85Kr volume activity in the atmosphere [4]. Sampling Discrete samples of atmospheric humidity were taken by freezing from sucked air in the condensation unit of our own construction. Sucked air was filtered with a fibrous filter AF PC (Slovak chemical works Hnúšťa) to prevent interference of tritium bound to aerosol. Before sampling started (the process of sampling usually lasted 6-8 hours), humidity in the monitored area had been determined with an aspiration humidity detector. In the case of cumulative samples a sample of atmospheric humidity was also frozen after previous filtration. The time of continuous sampling was usually a calendar month, during 1994-95 the interval of sampling was prolonged to three months. The 172 21st RHD Jasná pod Chopkom corresponding humidity for conversion to volume activity of the atmosphere was determined from the data obtained in meteorological stations of the Czech Hydrometeorological Institute in Praha-Libuš and Ruzyně. Activity measurement Tritium activity in obtained aqueous samples was measured with a liquid scintillation counter Tri-Carb 1050 (Canberra-Packard). For preparation of scintillation cocktail scintillators Instagel and Pico Fluor LLT of the same producer were used. The conditions of tritium activity measurement are described in detail in Ref. [5]. RESULTS AND DISCUSSION Results of tritium volume activity measurement in humidity in chosen areas are summarised in Table 1. Given are results of measurement from the year 1992, i.e. measurements from the last year of the generator operation, and those from the years 1995 and 1997. The chosen areas include in addition to the laboratories in which tritium targets were handled also the Institute's conference room, which could be contaminated only indirectly by particles from targets or by airflow in the building. The highest values were repeatedly measured in the radionuclide storage room in which both new and used targets were stored. The lowest values were paradoxically measured in the irradiation room in which the neutron generator was placed. We suppose that the mentioned values are related to the installed independent ventilation in this room and with general reconstruction of the room after the removal of the generator (including besides other things reflooring, new paintings, etc.) and also to accurate cleaning of the irradiation room during operation of the generator as well as in the later period. In the last column of the Table half-lives of tritium activity decrease are given. The half-lives are calculated assuming exponential decrease of activity and in comparison with tritium half-life (Tia = 12.3 years) they express the rate of tritium removal from the monitored areas. This parameter also confirms a marked decrease in contamination of rooms which were reconstructed after the removal of the neutron generator (e.g. irradiation room, isotope storage room). A higher value of half-life in the laboratory No.213 shows possible occurrence of target fragments in this room. In Fig. 1 results are given of long-term monitoring of tritium in outer atmosphere in the near vicinity of the Institute. From this figure a quick decrease of tritium activity in 1993 (after dismantling of the generator and liquidation of explored targets) is remarkable. The values of volume activity decreased from about 400 Bq.m'3 to the level of about 20 Bq.m"3 of the air. This value is however by three orders of magnitude higher than present background concentrations of tritium in ground level air of Prague. From the given results it is also obvious that even in the period of the highest contamination the limit of tritium concentration in the atmosphere valid that time for workers handling ionizing radiation was not exceeded, its value being 1.8 x 105 Bq.ni . The highest measured concentration in the radioisotope storage room, which was entered only occasionally, reached only 25 % of the mentioned limit. 2 ľ RHD Jasná pod Chopkom 173 References Kocol H., McNelis N.D., Moghiss A.A.: Health Phys.31: 76-78 (1976). Bárta K., Turek K.: Jaderná energie 18:347 (1972). Wilhelmová L., Tomášek M., Štukheil K: Tritium in Atmospheric Humidity in a Neutron Generator Facility (in Czech). Proc. 16th RHD Štrbské Pleso, p. 23 (1992). Tomášek M., Wilhelmová L.: J. Radioanal. Nucl. Chem. 218:119-121 (1992). Wilhelmová L., Tomášek M.: Jaderná energie 38: 405-408 (1992). Table 1. Contamination of chosen areas of the Department of Radiation Dosimetry by tritium (HTO) Place of sampling Volume activity [Bq.m"3] Tia June 1992 January 1995 November 1997 [y] Irradiation room of the 2.3 x 102 9 2.5 0.3 neutron generator No. 36 Radionuclide storage room 5.4 x 10" 6.5x1 C 3.2 x 10* 0.5 No. 03 Easily available radionuclide 8.4x10* 2.5 x 10' 67 0.9 storaqe room No. 7 Active laboratory No. 212 5x102 1.3 x102 25 0.8 Active laboratory No. 213 1.9 x102 95 15 1.4 Conference room No. 126 1.5 x102 50 5.5 0.9 Roof (outside the building) 2x102 50 20 0.8 1000 1990 1991 1992 1993 1994 1995 1996 1997 year Fig. 1: Volume activity of tritium (HTO) in the air in the vicinity of the building of the Department of Radiation Dosimetry. 174 SK98K0371 2ľRHD Jasná podChopkom CONTRIBUTION TO THE PENETRATION OF RADIONUCLIDES ACROSS THE SKDV. CONCENTRATION DEPENDENCE OF 60Co PERMEATION Zoltán Kassai, VasttKoprda, Margita Harangozó Chemickotechnologická fakulta STU, Radlinského 9, 812 37, Bratislava The health consequences of skin contamination with radioactive substances are radiation damage to the skin (local injury) on one hand, and possible secondary internal (systemic) uptake of radionuclides on the other [ 1 ]. Uptake of radionuclides by skin is fully different in case of intact skin, than it is in case of damaged skin, being from tens to hundred times higher in case of skin scarified, burned or wounded. Substances may enter (or leave) cells by: • passive diffusion (the rate is a function of the difference between the external and internal concentration of a substance), • facilited diffusion (characteristic saturation rate is achieved because the process depends on a finite supply of carrier molecules), • active transport (is characterized as an energy-requiring process, with saturation phenomenon, able to work against concentration gradient, showing a competitive inhibition), • pinocytosis (phagocytosis-like mechanism facilitating the absorption of large molecules). There exist three main routes of entry for substances into an organism through human (annual) stratum corneum (horny layer) that represents the major rate-limiting skin barrier: • Intercellular route (mostly lipid domain); • Transcellular route (mostly proteinaceous domain); • Appendageal route (hair follicles, sweat ducts, sebaceous glands). Radioactive cobalt belongs to frequent potential contaminants of human body. Its 60Co radionuclide belongs to the most dangerous components of the activated radionuclides in construction materials of nuclear piles. Aims • Using radioactive indicator 60Co to determine the speed of permeation of Co2+ across the skin depending on concentration of Co2+ in donor vehicle. • To research the dominant routes of radionuclide permeation (intercellular , transcellular or appendageal). • To assess the time profiles of Co2+ penetration across the skin. Experimental The experiments were done using double charged Co cation. The radionuclide fi0 Co was used with its chloride carrier CoCl2. Experimental arrangement consists of Franz-type vertical permeation cells [2] (active area 0.8 cm2), used with fresh skin from 2ŕ RHD Jasná pod Chopkom 175 abdominal region of 5 or 9 day old rats (5DR, 9DR resp.) of Wistar strain (Breeding Farm Dobrá Voda, SK). 5DR are still hairless and 9DR are just short haired. The 5DR skin was used either intact or with different thickness of horny layer after the skin had been stripped with adhesive tape (Scotch 3M) [3] 10 and 20 times, or the skin had been splitted (30 sec. in 60 DC water), whereby the whole epidermis was removed. Ions, penetrated through the skin from donor solution (vehicle water) to receptor solution (phosphate buffered saline, 1:9; pH 7.4), were determined in aliquots (0.3 ml) sampled in 1, 3, 5, 7 and 24 hours. The permeation cells were kept at 32 °C (surface temperature of the skin) during the experiment. The volume of donor compartment was 0.3 ml, the volume of receptor 7.3 ml. The concentrations of carrier in donor solutions were different (from 0.01 % to 1.0 %). The radionuclide 60Co was measured by means of Gamma Automat, Tesla, SK. Permeated fractions and fluxes were calculated regarding the sampled amounts by the homemade PC-program PERMEA. g 0,025 i 0,02 c=0,01% £ 0,015 - sn 0,01 - o —° c=0,1% 0,005 - íí- a 0 4 8 12 16 20 24 28 time (hours) Fig. I: Time profiles of permeated fractions of Co2+ across 10 times stripped skin at different donor carrier concentrations Fig. 2: Time profiles of permeated fractions of Co2+ across 20 times stripped skin at different donor carrier concentrations 176 21" RHD Jasná pod Chopkom Results and Discussion The results in Fig. 1 and Fig. 2 show that the penetrated amounts of Co2+ are proportional to time for at least in the first 7 hours. The results of permeation of ions across the intact, 10- and 20 times stripped skin at the same concentration showed that the more is the skin stripped, the more enhanced the penetration of ion is. The penetrated fraction was approximately 20 times higher in case of 20 times stripped skin than in 10 times stripped one. splitted 20x stripped 10x stripped & intact 12 16 28 time (hours) Fig.3: Time profiles of permeated fractions of Co2+across intact, 10-, 20 times stripped and splitted skin at donor carrier concentration c = 0.1% The results certify the fact that stratum corneum represents the most important barrier function of the skin. The penetration of Co2+ through the intact skin is less rapid than through 10-, 20 times stripped skin and yet far less rapid than in the case of splitted skin, as can be shown from Fig. 3. In the case of splitted skin the permeated fraction is roughly 300 times higher compared to the intact skin. Comparing penetration Co2+ ions through the 5DR skin (yet without hairs) with just haired skin (9DR), it was certified that the additional, transfollicular [3] shunt flux through the channels along the hairs (follicles) can be an important value comparing to transepidermal flux across the compact stratum corneum. The results in Fig. 4 showed that in the case of 9DR skin the permeated Co2+ fraction is about four times higher than in the case of 5DR intact skin. 8 12 16 28 time (hours) Fig.4: Permeated fractions of Co2+ through the skin with (9DR) and without (5DR) hairs at donor carrier concentration c = 0.1% 2ľ RHD Jasná pod Chopkom 177 0,7-, .. n 0,6- b * — ' C=0,01% 0,5- •P r •D 0,4 - 0,3 - / äeu u " *c=0,1% n 0,2- S. 0,1 - n c=0,5% CI 4 8 12 16 20 24 28 time (hours) Fig.5: Time profiles of permeated fractions of Co across splitted skin at different donor carrier concentrations 12 16 20 24 28 time (hours) Fig. 6: Time profiles of permeated fractions of Co2+ across the intact skin of 5DR at different donor carrier concentrations The results in Fig.5 and Fig. 6 showed, that at the lowest concentration of Co2+solution the permeated fraction is six times higher than at the higher concentration of cobalt ions. The lower is the concentration of Co2+ the higher is the permeated fraction (as the permeation fraction express the relative values) and on the other side, at these conditions, the lower is the concentration of Co2+ the lower are also the penetrated absolute amounts of substances. Conclusion Concluding it can be said, that: • the permeated fractions are the higher the lower is the concentration of Co2+ in donor solution, • the permeated amounts of Co2+ are proportional to the concentration of these ions in donor vehicle, • the dominant route of Co2+penetration is along the follicles, • it was proved that the principial penetration skin barrier for Co2+ is the horny layer. 178 21" RHD Jasná pod Chopkom References 1. J. Severa, J. Bár, Kontaminace radioaktivními látkami a dekontaminace, ČsKAE, ÚIS JP, Zbaslav, 1985 2. T. J. Franz: Percotaneous absorption on the relevance of in vitro data. J. Invest. Dermatol. 64, 190,1975 3. R. H. Guy, N. Higo, A. Naik at all, In: Prediction of Percutaneous Penetration, Vol. 2, (Eds.: R. C. Scott, R. H. Guy, J. Hadgraft, H. E. Boddé), p.l IBC Technical Services Ltd., London (1991) 4. J. Wepierre, O. Ducet, J-P. Marty: Percutaneous absorption of drugs in vitro: role of transepidermal and transfollicular routes. In: Prediction of Percutaneous Penetration, (Eds.: R. C Scott, R. H. Guy, J. Hadgraft), pp. 129-134, IBC Techical Services Ltd., London, (1990) 21st RHD Jasná pod Chopkom SK98K0372 179 FINDS OF RADIOACTIVE MATERIALS IN CENTRAL SLOVAKIA IN 1996 -1998 Ludmila Auxtová, František Ďurec, Pavol Adámek State Institute of Public Health, Cesta k nemocnici 1, 975 56 Banská Bystrica, Slovak Republic Abstract A brief report of finds of radioactive materials out of control in Central Slovakia during the period 1996 - 1998 has been made in this paper. The reasons of the reported events are described if known and some recommendations on safety culture development and implementation have been given. Background The application of radiation sources in industry, agriculture, research, teaching and medicine is broad and wide-spread in Slovakia. Political and economical changes after 1989 resulted in changes in organisations and their responsibilities as well as in ownership of factories and companies.The more or less qualified staff and most senior workers were often dismissed by the new owners. The new employees - untrained persons and lack of documentation have led to increase of probability of losses and uncontrolled movement of sources of ionizing radiation and radioactive materials. Increasing prices for the disposal have also led to throwing the sources away instead of proper disposing by the user. Sources in such cases can finish up in small scrap yards, smelting works, industrial waste disposals, waste treatment plants and steel companies. Unexpected finds of radioactive materials The staff of Radiation Protection Department of State Institute of Public Health in Banská Bystrica was during the period 1996 - 1998 involved in 11 remedial actions related to unexpected finds of radioactive material in scrap-metal. Most of reported finds were located at the entrance to a steel company in Podbrezová or at the Podbrezová railway station, where the incoming railway wagons were monitored by the personnel of the steel company using of small dose rate meters. Now the steel company is equipped by an automatic monitoring system. One railway waggon was returned from Italy to engineering works in Považská Bystrica. From 50 000 kg of steel swarf more than 10 000 kg were contaminated by ^Co. 10 ^Co sources with containers were found during a regular inspection visit in a construction company, where they were prepared for transport with other scrap-metal to the steel producing company in Košice. The only reason they were still in the company was heavy rain causing difficulties in loading, they were found in fact in the last minute before accidental melting. The very last cases were 2 trucks returned from an Austrian company working with metal scrap as contaminated back to the firm in Žiar nad Hronom which has exported the aluminium scrap. The bags with aluminium scrap were more or less contaminated by an accidentaly melted 90Sr source from icing thickness mesurement devices used in the military air force. 180 21st RHD Jasná pod Chopkom List of unusual events and finds in Central Slovakia in last 3 years No Date Locality Material found Notes 1 Oct. 16, 1996 Podbrezová 3 open containers with There was no licence railway station 60Co sources issued for this sources, the owner was not found. 2 March 3,1997 Podbrezová 5 truck axles from steel railway station contaminated by 60Co 3 May 10,1997 Podbrezová Steel contaminated Part of an agricultural railway station by60Co machine 4 July 8,1997 Liptovský 10 containers with 60Co Found during an regular Mikuláš, sources inspection in „last minute" construction before loading to a railway company waggon together with other scrap destined for a steel producing company* 5 Sept. 20,1997 Podbrezová 6UCo source without The owner was found railway station container 6 Dec. 14,1997 Podbrezová, steel Steel contaminated Part of an agricultural 8,30 producing by60Co machine company 7 Dec. 14,1997 Podbrezová Steel contaminated Part of an agricultural 18,00 railway station by60Co machine 8 March 23, Považská Bystrica Swarf from machining Railway wagon returned 1998 engineering steel parts contaminated from Italy owing to higher works by60Co dose rate 9 July 16,1998 Podbrezová Metal scrap with "6Ra Parts of an ore grinder, deposition metal scrap from Romania 10 July 23,1998 Podbrezová Metal scrap with ^"Ra Parts of an ore grinder, deposition metal scrap from Romania 11 Oct. 21,1998 Žiar nad Hronom Aluminium scrap 2 trucks returned from an contaminated by 90Sr, Austrian factory probably of military origin - melted device for icing thickness measurement *The steel producing company was not equipped with a fix portal and there was no dosimetry entrance control system installed. Remedial actions taken In all cases where sources were found they were removed from the railway waggons and transported by personnel of the specialized organization HUMA LAB APEKO to the labs for analysis and identification. In other cases the radioactive material removed from the waggons or trucks was declared as radioactive waste and will be stored as such. All finds of parts of agricultural machines were of the same origin and so follow-up inspections were planned in agricultural companies to find and collect the contaminated parts of machines and spare parts. 21" RHD Jasná pod Chopkom 181 Results of follow-up inspections in agriculture cooperatives, service companies and companies distributing spare parts for agriculture machines: 1997 80 contaminated parts of machines found by inspectors from Radiation Protection Department of SIPH Banská Bystrica The found and collected parts were returned to the producer of agricultural machines in the Czech Republic and disposed on his storage place. 1997 more than 40 contaminated parts up to now were found with help of Regional and District Offices for Civil Protection Primary causes to the finds reported Event 1: „Orphan" sources Events 2. 8: Probably an accidentally melted source of unknown origin Events 3. 6. 7 : A spent 60Co radiotherapy source stolen from a hospital was accidentaly melted in a small company in the Czech Republic. Parts of agricultural machines were produced from the contaminated steel, with the activity about 3,5 MBq/kg. Events 4. 5: Inadequate training and failure to follow requirements of regulatory authority Events 9.10: Metal scrap was contaminated by NORM (Naturally Occuring Radioactive Material). Event 11: Inadequate education and training of personnel working with sources used in icing density measuring devices for air force. Consequences The exposure of people involved was in events reported fortunately very low, but mishandling of sources can result in the future in overexposure of members of the public. Conclusions There is an urgent need: to disseminate more information, the knowledge gained and lessons learned from the finds and losses and detailed circumstances of each case to people working in scrap yards and melting factories; to inform the organizations dealing with scrap materials about the potential existence of orphan sources and contaminated scrap of various origin. 182 21" RHD Jasná pod Chopkom It is very important to have relevant risk information and to communicate relevant information about the events to the public. It is necessary: to oblige (by act or decree ) the companies working with metal scrap to realize radiation measurements before transport - to improve training and qualification of individuals to implement radiation measurement systems at entrance of scrap yards. There is also urgent need for better cooperation between authorities responsible for solving such events aimed to find the possibility how to store the materials found and declared as radioactive waste. 2ľ RHD Jasná pod Chopkom S K98K03 73 j g3 AIRBORNE GAMMA RAY SPECTROMETRY - METHODS, DATA PROCESSING AND APPLICATIONS Klusoň J.1}, Cechák T.1}, Juna P. 2) !> CTU, Faculty of Nuclear Sciences and Physical Engineering, Dept. ofDosimetry and Application of Ionising Radiation, Břehová 7, 115 19, Praha 1, Czech Republic 2> PICODASPraha, s.r.o., Geologická 2,152 00, Praha 5, Czech Republic The main advantage of the airborne gamma ray spectrometry is ability to collect the requested data effectively and quickly even over the large areas. It predetermines this method for the monitoring and mapping of the distributions of natural as well as man-made radionuclides concentrations on the ground surface or in the soil surface layer. Geological prospection and mapping, environmental and radioactive contamination studies, operational and accidental monitoring of the nuclear facilities neighbourhood, etc., are typical examples of application. Due to requirement to measure even low activities (on the natural background level) in distant geometry (usual flight altitudes are from about 40 up to 250 meters), the high sensitive detection systems are necessary. Most of systems use large volume NaI(Tl) detector blocks. To derive quantitative information from the spectrometry data, the calibration of such detection system is necessary. There are two ways to calibrate the airborne spectrometer: - experimental, using calibrations pads with known selected radionuclides concentrations and/or flights over the calibrations fields with known concentrations, found by the ground measurements or by sampling - theoretical, using stochastic modelling (by Monte Carlo method) of the detector responses, corresponded to the unit concentrations of the considered radionuclides While experimental method is limited by the selection of the available sources and experimental arrangements, the theoretical approach makes it possible to simulate any source type, configuration and depth distribution as well as source - detector geometry. The method of the airborne spectrometry data mathematical processing based on the model calculations of the detector response/response matrix was designed and its application for the ENMOS airborne gamma-ray spectrometry system (produced by PICODAS Group. Inc.) was developed [1,2]. Using the prepared Monte Carlo code, the spectrometer responses for natural radioelements (^K, U and Th-series), selected man-made contaminants (including depth distribution models) and set of flight altitudes were calculated. In the first step, the photon field angular energy distributions for all considered source configuration models and geometrical arrangements were simulated. Example of this distribution for the Th-series and altitude 100 meters is in the Fig. 1. In the second step, the detection system responses for fields with calculated characteristics were simulated. Example of result for U-series and flight altitudes 60, and 100 meters is in the Fig. 2. Mathematical methods for the experimental spectra decomposition to the individual component (radionuclides) contributions (based on calculated responses and least squares method or the deconvolution technique) was developed for the spectrometry data processing and evaluation. Developed method was widely tested and successfully used in number of applications, e. g. for environmental studies [2,3]. Models, used for theoretical response calculations are usually based on some approximations and/or simplifications and assumptions, describing real experimental arrangement. For instance the real terrain/source is approximated by the semiinfinite. planar soil-air interface with defined source depth distribution in the soil surface layer, constant mean soil/air density and humidity is considered. Description of the complex detection system is usually simplified as well as inclusion of the detector shielding by the plane, etc. Correct interpretation of results then depends on the variations of the real experimental conditions 184 21" RHD Jasná pod Chopkom 0.14- •o 0.12-:' i 0.1-....;; cos-;::-- éCm 0.06- '['...•- "O 0.04-:::::' «*• •o 2000 ío 25000 EfkeV] Fig. 1 Calculated ENMOS Detector Response Functions for U-series 0.35 1 60 meters 1 100 meters 0.3 0.25 \ i I 0.2 0.15 0.1 " A A K ... 0.05 A 0.5 1.5 2.5 Energy [MeV] Fig. 2 comparing them with assumptions involved in the models. Affect of the soil density, real terrain profile and roughness, attenuation by vegetation, etc. should be taken into account. For these purposes, the studies of the spectrometer response dependence on the variations of discussed factors were done using prepared models [4]. Relative response dependence on the soil density calculated for natural radionuclides and 137Cs (for altitude 100 meters) is in the Fig. 3. 21aRHD Jasná podChopkom 185 Calculated corrections for the vegetation cover superficial density for selected energy lines of natural radioelemets as well as for 137Cs (with exponential depth distribution) and altitude 100 meters are in the Fig. 4. Different character of the dependence (nearly independence) for the 137Cs (Fig. 3) results from different (exponential) depth distribution. Vegetation cover type and density (soil humidity) can be estimated from multispectral satellite images of given region [3]. 1.3 100 meters 1.25 137Cs * s I 1.2 U-series o 3 A •e 1.15 Th-series 1.1 .....JSfe., 1.05 x 1 0.95 0.9 v DC 0.85 - N... -^ 0.B 1.3 1.4 1.5 1.6 1.7 1,8 1.9 2.1 Soil density [gem'3] Fig. 3 2.2 Altitude 100 meters 2 I>83 keV (208TI) o 2 - 462 keV ŕ°K) A 2£i15keV(2DBTI) o 6 52 keV (137Cs) * 1.6 o 1.4 - 1.2 10 Vegetation cover superficial density [g cm"' Fig. 4 Next aspect of the measurement interpretation is averaging of data over the large area, when weight of the contribution is highest for the central circular area under detector and 186 2 ľ' RHD Jasná pod Chopkom decrease for more distant regions. The spatial resolution depending on data collection method parameters (aircraft velocity, flight altitude and time of one spectra acquisition - i.e. one scan) can be introduced. Dependence of relative response on the central circular region radius was calculated for considered radionuclides and set of flight altitudes. Example of results for the 137Cs and altitudes 40, 60, 80, 100, 120 and 200 meters in the form of the areas of central circle and concentric circular rings corresponding to contributions 0-25%, 25-50%, 50-75% and 75-95% are in the graph and table, see Fig. 5. Knowledge of the measurement spatial distribution can be useful for interpretation measurements on the broken terrain, comparison results with results from the ground measurements, etc. ~E 600000- flight altitude |m] Fig. 5 It can be concluded, that method under discussion is successfully used and that new, computational model based analysis of factors, affected the result interpretation, enables better uncertainties assessment and/or appropriate corrections design and evaluation. [1] ČECHÁK, T. - KLUSOŇ, J. - MALUŠEK, A.: Spectra Processing in Airborne Gamma- Ray Spectrometry Using the ENMOS Detection System, Research Report, Contract No. 402994, KDAIZ ČVUT FJFI Prague, May 1994. [2] JURZA, P. - PICHL, R. - KLUSOŇ, J. - ČECHÁK, T. - DRÁBOVÁ, D.- FILGAS, R.- ČEŠPÍROVÁ, I. - MATOLÍN, M.: Využiti moderních metod pro hodnocení zátěže životního prostředí se zaměřením na vliv těžby uranu, zpráva projektu Va V/630/3/97 MŽP ČR, Praha 1997. [3] DRÁBOVÁ, D. - FILGAS, R. - MALÁTOVÁ, I. - ČEŠPÍROVÁ, I. - FOLTÁNOVÁ, Š. - HÔSCHL, V. - ROUDNÝ, R. - JURZA, P. - PICHL, R.: Survey of post-Chernobyl contamination in Šumava region using various methods. Proceedings of the IRPA regional symposium on radiation protection in neighbouring countries of Central Europe, September 8-12, 1997, Prague, Czech Republic. [4] KLUSOŇ, J. - ČECHÁK, T.: Some Aspects of the Airborne Gamma Spectrometry Data Processing and Interpretation, Acta Polytechnica (submitted for publication). 21" RHD Jasná pod Chopkom SK98K0374 187 A COMPARISON OF SEMICONDUCTOR GAMMA SPECTROMETRIC ANALYSIS USING THE PEAK NET AREA CALCULATIONS AND THE WHOLE SPECTRUM PROCESSING Š. Krnáč1}, M. Koskelo 2), R. Venkatamaran 2> 1} Department of the Nuclear Physics and Technology, Slovak Technical University, Ilkovičova 3, 81219 Bratislava, Slovakia 2) Canberra Industriesjnc, 800 Research Parkway, Meriden, Connecticut 0645, USA 1. Introduction A study was conducted to compare the results of gamma spectrometric analysis using the SCFA (Scaling Confirmatory Factor Analysis) method to that of Genie2K, which uses a more traditional method. The SCFA method is based on whole spectrum analysis whereas the traditional method is based on the quantification of the net peak area of the full energy absorption peaks. The SCFA method has been proposed by [1,2]. In applying his method, a 'significant factor structure' has been extracted for a given source-detector geometry, at a given energy. The 'structure' includes, (1) a factor for total absorption (interactions ending with the photo-electric effect), (2) a factor for Compton scattering (primary and secondary), (3) a factor for backscattering (mainly from the top and side walls of the shield), and (4) a factor for low and high energy noise associated with charge collection. The functional form of each one of the factors has been proposed with a good fit at 10 different standard gamma ray energies. 25 parameters of the factor functions have been determined at the standard energies using least squares minimization. Three sets of parameters have been determined, one for each geometry, using which the Standard Response Operator (SRO) has been computed for each geometry. The Complex Response Operators (CRO) are obtained by interpolation between the standard parameters, over the entire energy range of interest. Interpolation functions have been proposed for each standard parameter. The interpolation step allows the construction of a response matrix for a given source-detector geometry [1]. 2. Spectral analysis using Genie2K Gamma ray spectra had been acquired for several gamma standard sources, all of which except Co-57 and Eu-152 being single gamma ray emitting nuclides. These standard sources spanned the energy range from 60 keV (Am-241) to 1116 keV (Zn- 65). The standard sources were counted at 3 different geometries, with source-detector distances of 0, 5, and 15 cm. A certificate file was created by combining the activity information of all the standard sources. Using single gamma ray spectra collected at a given counting geometry, and the certificate file, an efficiency calibration was created for that geometry. Three different test spectra, one for each counting geometry, had been created by combining several of the standard source spectra. The efficiency calibrations created for the 3 geometries were loaded into the respective spectrum files. Each test spectrum was analyzed using the standard Genie2K engines; Peak locate, Peak search. Interactive peak fit, Background subtraction, Efficiency correction, and Nuclide Identification with interference correction. The results of the various calculation steps were reported. In the case of Test spectrum 2, an additional reporting step was run using the template 188 2 ľ RHD Jasná pod Chopkom Area.TPL. This additional report gives the % contribution to a peak area from a given nuclide, and therefore, is useful when analyzing spectra that consist of interfering gamma ray lines (for example the 122 keV line from Co-57 and Eu-152). The interference corrected nuclide activities from each test spectrum were compared to the corresponding activity results provided by the SCFA method. In each of the three test cases, the SCFA method correctly identified the nuclides that were present in the gamma ray spectrum. The results of quantitative comparisons of activities are given in Tab. 1, 2, 3, and 4. The results of the comparison are briefly summarized in the following paragraphs. 3. Comparison of SCFA and Genie2K activity results with the Reference activity TESTI: Test spectrum 1 was created by combining the standard source spectra from Am- 241, Cs-137, and Te-123m, obtained at the counting position of 5 cm. The background spectrum was included in the test spectrum. The ratio of activities, Genie2K / Ref, and SCFA / Ref were determined. An unweighted average and standard deviation were determined for each set of ratios. The mean values of the both ratios, Genie2K/Ref and SCFA/Ref, overlapped unity within 1 sigma standard deviation. Genie2K ratios were more tightly distributed than the SCFA ratios. Energy Activity Activity Reference Nuclide G2K/Ref SCFA/Ref (keV) (Bq) (Bq) (Bq) G2K SCFA Am-241 59.54 39113.4 39770.04 39240 0.997 1.014 Te- 159 24808.44 23891.52 24720 1.004 0.966 123m Cs-137 662 39578.88 38474.76 39720 0.996 0.969 Unwtd. 0.999 0.983 Avg. Std. devn. 0.004 0.027 Tab. 1 Results of comparison for test spectrum No. 1 TEST 2: Test spectrum 2 was created by combining the background spectrum with standard source spectra from Zn-65, Cd-109, Co-57, and Eu-152, obtained at the counting position of 15 cm. This spectrum was interesting in two respects. First of all, the interference at the 122 keV line due to Co-57 and Eu-152 needed to be sorted out. Secondly, a potential interference at 88 keV from the nuclides Pb-214 and Pb-212 needed to be accounted for. The physical semiconductor gamma spectrum 2 is plotted in Fig. 1. The unfolded emission rate spectrum 2 is depicted in Fig. 2. An unweighted average of the activity ratios at all other energies indicated that Genie2K had a positive bias of 10%, while, the SCFA showed a bias of only +2.2%. The SCFA ratios were more tightly distributed with a 1 sigma standard deviation of ±0.044, whereas the Genie2K ratios had a standard deviation of ±0.149. The results corresponding to gamma lines suspected of interference were excluded from the average and the standard deviation. 2 ľ RHD Jasná pod Chopkom 189 ; : ŕs !l!fi§ i : ;i - -,. V :,X;- : r-. A": t.; ?• •; - i"!.- - "Í i/-' |||ä : t IBIII )•:.'• j- .-•'.•!•• "'ľ-•;;-•> ••'*•':"•• "•'••-J :'ŕ-'•]•" v 1 ';:'--^"-.|;-:-;-v;i"1vi;:. pil; '".•.:;.;:;::':.^í'lii. ^^rí-A-f^vH 1 0.1 Wimk • ;!• i~r- ••-,;'.-.*.>•'; ••,•'.>' •* • -í'•'•'•• '••'•> i • ' '• -.',t':!•'••!-. ••'•' •-"T^j"•-];•'•!•';' 0.01 life •••";:-.: "•: ;/;• •:'.•'->'••' J šili Hli 0.001 0.0001 1 1001 2001 30O1 Ctunm) Fig. I Physical test spectrum No.2 ľ,' ' í".' 778 10B1 1388 1512 1757 Fig. 2 Emission rate spectrum No.2 To its credit, the SCFA method identified a number of low yield gamma ray lines from Eu-152, which were not identified by the Genie2K analysis. Further, the SCFA method came up with reasonable activities for these low yield Eu-152 gamma lines. The mean of the ratio of these activities with respect to the reference was 1.112, with a standard deviation of ±0.174. 190 21" RHD Jasná pod Chopkom Energy Activity (Bq) Activity (Bq) Reference Nuclide G2K/Ref SCFA/Ref (keV) G2K SCFA (Bq) Cd-109 88.03 832.71 1218.50.5 1448.92 0.575 0.841 Co-57 122.06 2489.04 2121.98 2060.92 1.208 1.029 Co-57 136.48 2548.47 1729.44 2060.92 1.237 0.839 Zn-65 1115.52 1053.42 845.13 844.56 1.247 1.001 Eu-152 121.78 1997.61 1893.93 1853.85 1.078 1.021 244.69 2019.75 1927.47 1853.85 1.089 [.040 344.27 2000.61 1865.55 1853.85 1.079 .006 411.11 2252.43 1852.14 1853.85 1.215 ().999 443.98 1934.46 1959.99 1853.85 1.043 .057 778.89 2024.97 2020.26 1853.85 1.092 .090 867.32 1458.72 2013.24 1853.85 0.787 .086 964.01 2451.69 2005.17 1853.85 1.322 .082 1085.78 1850.19 1966.14 1853.85 0.998 .061 1112.02 2007.33 1938.06 1853.85 1.083 .045 1407.95 1946.34 1784.28 1853.85 1.050 0.962 Unwtd. Avg. 1.104 1.022 Std. devn. 0.149' 0.044 Tab. 2 Results of comparison for test spectrum No. 2 Energy (keV) Yield (%) Activity (Bq^ Reference (Bq) SCFA/Reference 367.78 0.83 1971.75 1853.85 1.064 416.05 0.11 2416.2 1853.85 1.303 488.66 0.407 2040.06 1853.85 1.100 564.02 0.468 2011.98 1853.85 1.085 566.42 0.1294 1578.99 1853.85 0.852 586.29 0.459 1958.55 1853.85 1.056 656.48 0.144 1530 1853.85 0.825 674.67 0.167 2561.22 1853.85 1.382 678.59 0.461 2262.84 1853.85 1.221 688.67 0.835 1820.16 1853.85 0.982 841.58 0.1628 2230.08 1853.85 1.203 919.4 0.436 2382.9 1853.85 1.285 1005.27 0.647 2508.48 1853.85 1.353 1089.7 1.71 1890.33 1853.85 1.020 1109.18 0.183 2058.75 1853.85 1.111 1212.95 1.399 1755.57 1853.85 0.947 Unwtd. Avg. 1.112 Std. devn. 0.174 Tab. 3 Eu-152 lines identified by SCFA, but not by G2K TEST 3: Test spectrum 3 was created by combining the background spectrum with the standard spectra from Te-123m, Sn-113, Cs-137, Mn-54, and! Zn-65, obtained at the counting position of 0 cm. Once again, the SCFA/ Ref ratios were tightly distributed about a mean value of 0.987, with a standard deviation of ±0.015. The Genie2K to reference activity ratios had a mean of 1.198, with a standard deviation of ±0.252. The Genie2K results for Sn-113 and Zn-65 nuclide activities showed large positive biases. It 2 ŕ RHD Jasná pod Chopkom 191 needs to be pointed out that in both Test 2 and Test 3, the Zn-65 activity predicted by Genie2Kwas 20% higher than the reference activity. Energy Activity Activity Reference Nuclide G2K/Ref SCFA/Ref (keV) (Bq) (Bq) (Bq) G2K SCFA Te- 159 24684.6 24209.1 24440 1.010 0.991 123m Sn-113 391.62 208.2 128.1 130.65 1.594 0.980 Cs-137 662 169.5 169.2 170.796 0.992 0.991 Mn-54 834.96 6750.9 5754.9 5732.3 1.178 1.004 Zn-65 1115.78 2926.5 2324.7 2401.63 1.219 0.968 Unwtd. 1.198 0.987 Avg. Stá. devn. 0.252 0.015 Tab. 4 Results of comparison for test spectrum No. 3 4. Conclusion In all 3 test cases, the SCFA method identified all the nuclides correctly. The K- 40 activities calculated by the SCFA method were reasonably close to that from Genie2K analysis. In general, the quantitative results of the SCFA method were impressive in all 3 cases. On a positive note, the SCFA method did identify low yield gamma lines in Eu- 152, which were not identified by the Genie2K analysis. This substantiates claim that the SCFA is more sensitive than the traditional method of spectrum analysis. 5. References [ 1 ] Š. Krnáč, Gamma spectrometry as a latent problem, PhD thesis, MFF UK, Bratislava (1993). [2] Š. Krnáč, P. P. Povinec, J. Radioanal. Nucl. Chem., Vol. 204, No. 1, 57-74 (1996). [3] Š. Krnáč, P. Ragan, Radiat. Protect. Dosim., Vol. 58, No. 3,217-228 (1995). 192 SK98K0375 21st RHD Jasná podChopkom PLASTIC SCINTILATOR GUARD GATES FOR MONITORING THE RADIOACTIVITY OF THE RAILWAY WAGONS Štefan Krnáč Department of the Nuclear Physics and Technology, Slovak Technical University, Ilkovičova 3, 812 19 Bratislava, Slovakia 1. Introduction The reprocessing of waste materials is becoming increasingly important. At the same time, there is a growing risk that metal scrap, for example, may contain radioactively contaminated objects or even radiation sources. Only prompt detection of these components will ensure protection from complication and the resultant costs. The intentional and unintetional carrying of ilegal radiation sources in luggage or in vehicles also involves incalculable hazards. The increasing numbers of reports in the media indicate a threatening trend. 2. Monitoring system A new system of guard gates for radioactivity monitoring of the railway truks has been at present proposed. The device has been certified by the Slovak Institute of Metrology (see Refs. [1, 4]). A block scheme of the system is outlined in Fig. 1. The system includes the following main parts: i) Radiation detection (DETECTOR).The system is based on detection using the large area plastic scintilation detectors. Due to a high sensitivity of the detectors, two weather-proof plastic scintilation detectors react rapidly to any short-term rise in radioactivity and detect concealed components, even under tons of steel scrap. Also, multidetector system is available to use. ii) Wagons identification (SENSOR). By incorporating an external sensor signal (proximity sensor or weighing signal), the metering systems operate with maximum sensitivity and minimum false alarm rate. In order to read and store the identification number of wagon, a video system including the infra camera and video recorder may be employed. iii) Time analysis (FILTER). An in-house developed evaluation mode (ADF- Advanced Digital Filter) permits optimum measured value computation ideal for all velocities in a 250 milisecond cycle. An arithmetical evaluation process makes it possible to analyze and take into account the shielding effect of the vehicle cargo in relation to the natural background radiation. The alarm threshold is accordingly adjusted. Therefore, in many cases, sensitivity of measurement is in fact increased significantly. iv) Spectral analysis (MCA). When the limit value is exceeded, contamination alarm is triggered and spectral analysis is started. The measurement signals are fed into a multichannel analyzer and plastic scintilation spectra are obtained. The spectra are output with date and time of day via interface for storage, documentation and next evaluation analysis. v) Controling and synchronisation (CONTROL). The signals and data from the wagons identification sensors, alarm settings and spectral analysis are synchronized and used 21" RHD Jasná pod Chopkom 193 for controling all processes of the monitoring. The PC-compatible computer ensures the necessary flexibility in data processing with a high level of transparency. Data transfer to a host computer is effected via V.24 interface which also permits remote control of the unit. vi) Evaluation and certification (METRA). By setting alarm thresholds, it is possible to trip alarms and initiate logical decision. The various options of internal measured value processing are freely programmable. For example, the METRA software enables to convert plastic scintilation spectra to photon flux energy distributions and, accordingly, to calculate exposure and/or dose quantities in order to evaluate correctly the alarm state being occured. The process is based on the response operator evaluation of the spectra (see Refš. [1,2,3]). DETECTOR SENSOR Plastic scintilators Wagons counter Photomultiplier tube Video identification High voltage supply Storing video records JTER CONTROL Time analysis Controling processes OUTPUT Alarm setting with Synchronisation background Automatisation MCA METRA Spectra acquisition Spectra coversion OUTPUT Spectral analysis Evaluation of dose rates Storing alarm spectra Certification and storage ••^^ from measurement signals to measured and evaluation values, *• controling and synchronisation signals Fig. 1 Block scheme of the monitoring system 3. Working of the guard gate The scintilation guard gate has been inslalled at the railway cargo station in Bratislava-East. The 24-hour-working of the monitoring system of the guard gates has been proven by tests within several days (from Oct., 30 up to Nov., 3 1997). In this time, the number of gated railway wagons achieved approximately 11,000 objects being monitored. The system worked properly all the time. The alarm threshold was set at 30 % over natural background. Average counts rate from the natural background was about 2600 cps, it approximately corresponds to dose rate in air of 92 nGy/h. Minimum values of the adjusted background under the shielding effect of the truck and the cargo were 2076 cps and 58 nGy/h. During the working, only 51 cases axceeded the setting alarm threshold. The calculated dose rates in air (exposure rates) for these alarm cases are depicted in Fig. 2. Protocol of the testing is introduced in Tab. 1 where number of the alarm, counts per second, photon dose rates in air and time identification number of the alarm measurement are listed for all alarm wagons. 194 21st RHD Jasná pod Chopkom 400 I 350 ' . 300 • 250 200 150 100 ; WE ft] 50 •! 1 4 7 10 13 16 19 22 25 28 31 34 37 40 43 46 49 Man bar of alarm Fig. 2 Photon dose rates in alarm cases In respect of the results, it may be found that alarm values of the dose rates do not exceed 160 nGy/h except for the alarms No. 13, 14, and 39 (266.6 nGy/h, 398.0 nGy/h, and 242.9 nGy/h). 4. Spectra! analysis of the alarms In order to determine the possible source that increases the measured dose rates above the limit value, the spectral analysis of the alarm spectra may be a promising way. The two 128-channel physical alarm spectrum of No. 14 and 39, respectively, gathered from the plastic scintilation detectors are plotted in Fig. 3. These physical wide-spread spectra are unsuitanle for qualitative analysis of radiation. However, after their conversion to incident spectra (photon flux and/or dose rate energy distribution) using the response operator method [2,3], the situation is significantly changed. 1000 \ •o i 100 i 10 Count s 1 11 21 31 41 51 61 71 81 81 101 111 121 Channel Fig. 3 Physical plastic scintilation spectra for the alarm No. 14 and 39 2ŕ RHD Jasná pod Chopkom 195 COUNT STD RSD DOSE STD RSD NAME No. (cps) (cent) (nGy/h) (nGy/h) (cent) (hhmmssdd.mm) 1 4061.8 +/- 456.2 11.2 145.8 +/- 17.9 12.3 8472003.11 2 4101.6 +/- 444.1 10.8 144.7 +/- 17.3 11.9 8471403.11 3 4019.2 +/- 466.2 11.6 149 +/- 18.8 12.6 8470803.11 4 4113.2 +/- 356.6 8.7 153.1 +/- 15.3 10 8470303.11 5. 4050 +/- 376.8 9.3 150.4 +/- 15.9 10.6 8465203.11 6 4137.2 +/- 453.1 11 159.5 +/- 19.2 12 8464603.11 7 3960.6 +/- 425.8 10.8 147.6 +/- 17.5 11.9 8464003.11 g 4016.6 +/- 369.1 9.2 146.3 +/- 15.3 10.5 8463403.11 9 3867 +/- 330.1 8.5 143 +/- 14.1 9.9 8462903.11 10 3957.2 +/- 409.7 10.4 142.9 +/- 16.4 11.5 8462303.11 11 3782.6 +/- 373.8 9.9 141.7 +/- 15.7 11.1 8461703.11 12 2501.4 +/- 286 11.4 88.3 +/- 11 12.5 6152003.11 13 6656.6 +/- 401.1 6 266.6 +/- 20.9 7.8 6151403.11 14 9996.2 +/- 607.1 6.1 398 +/- 31.3 7.9 6150803.11 15 3291.8 +/- 269.5 8.2 127.3 +/- 12.2 9.6 6150303.11 16 3982 +/- 354 8.9 146.8 +/- 15 10.2 17J\3950.4 422.5 10.7 147.7 +/- 17.4 11.8 ^^^^ w ^51240vH -ft oi-y/f 34 4318 482.7 11.2 158.1 +/- 19.4 12.2 8344301.11 35 3564.4 +/- 368.3 10.3 125.7 +/- 14.4 11.5 8343801.11 36 4421.4 +/- 395.8 9 159.1 +/- 16.3 10.3 8342101.11 37 2492.2 +/- 315.8 12.7 86.5 +/- 11.8 13.6 1292231.10 38 3612 +/- 374.2 10.4 129.5 +/- 14.9 11.5 1291731.10 39 6275 +/- 513.4 8.2 242.9 +/- 23.3 9.6 1291131.10 40 3794.6 +/- 439.2 11.6 144.2 +/- 18.2 12.6 1290531.10 41 3765.8 +/- 345.5 9.2 135.5 +/- 14.2 10.4 8132330.10 42 3642.4 +/- 338.7 9.3 134.9 +/- 14.2 10.6 8121030.10 43 3948.6 +/- 449.4 11.4 137.2 +/- 17.1 12.4 8120430.10 44 4043.8 +/- 374.8 9.3 152.1 +/- 16 10.5 8115930.10 45 3777.6 +/- 399.9 10.6 139.7 +/- 16.4 11.7 8115330.10 46 3714.6 +/- 389.5 10.5 143.7 +/- 16.7 11.6 8114730.10 47 3998.2 +/- 489.4 12.2 139.7 +/- 18.5 13.2 6171830.10 48 3849.8 +/- 393.4 10.2 136.1 +/- 15.5 11.4 6171230.10 49 2959.4 +/- 327.8 11.1 104.4 +/- 12.7 12.2 6170630.10 50 3929.6 +/- 437.7 11.1 137.6 +/- 16.8 12.2 6170130.10 51 4658.6 +/- 397.9 8.5 161.6 +/- 16 9.9 5044230.10 Tab. I The 5-days alarm protocol of the wagon guard gate An example of the converted photon dose rate spectra for the alarms No. 14 and 39 is shown in Fig. 4. The depicted values represent the photon dose rates in air per energy interval of 14 keV. In the both cases, it may seen that a great deal of dose contribution is originated from a region about 1460 keV, which corresponds to the radiation source of 40K. A similar considerable 40K contribution to the total dose can be found also within the other alarm spectra. 196 21" RHD Jasná pod Chopkom Fíg. 4 Photon dose rate 14-keV-distribution for the alarm No. 14 and 39 5. References [1] Š. Krnáč, In Situ Gamma Ray Counting of the Large Volume Objects Using the Large Area Plastic Scintilators, in this Proceedings (1998). [2] Š. Krná5, P. P. Povinec, J. Radioanal. Nucl. Chem., VoL 204, No. 1, 57-74 (1996). [3] Š. Krnáč, P. Ragan, Radiat. Protect. Dosim., Vol. 58, No. 3,217-228 (1995). [4] Certificate of Metrological Verification No. 007 250 2192/97, Slovak Institute of Metrology (1997). 21" RHD Jasná pod Chopkom SK98K0376 197 THE UNINTENTIONAL AND UNCONSCIOUS EXPOSURE TO RADON (AND OTHER NATURAL RADIONUCLIDES) Josef Thomas State Office for Nuclear Safety, Senovážné nám. 9, Praha 1 Introduction Beside of commonly applied radiation protection to practices, where natural radiation sources (NRS) are utilised intentionally and consciously (mining of uranium, radiotherapy by radium or radon etc), and where the principles of radiation protection are fully applied, the approach to exposures from NRS obtained unintentionally and unconsciously is considerably different and more reserved, although such exposures are more frequent and mostly higher then exposures from NRS obtained intentionally and consciously. There are two types of such exposures, first the exposure of the population by radon in their homes, where the responsibility is on the houseowner, second the exposure to radon (and to other natural radionuclides) in common workplaces, as offices, workshops, underground workplaces etc and in workplaces processing natural raw materials with enhanced levels of NRS and producing radioactive residues, where the responsibility is on the employer. These two cases are, seen from the point of view of law, markedly different and need different approaches of the radiation protection. Both cases are treated by radiation protection as interventions against inadequate and unacceptable high exposures and not as licensing or controlling of practices (leading to exposure), according to international recommendation, e.g. the IBSS [1] or the Czech Atomic Act [2] and the providing decree [3]. In such interventions it is emphasised always, that interventions can be applied (are justified) only against such contributions from the NRS, which are controllable by technical countermeasures or by changes in the utilisation and/or of the regime and which are significantly higher above the background, e.g. against higher levels of indoor radon, against higher levels of NRS in building materials and in water and not against any levels of radon, against the content of NRS in the body or in food (if not caused by human activities), against the terrestrial and cosmic radiation (except for air crew) etc. Radon Programmes for dwellings a) Remediation of the risk The case of exposure mentioned first is treated in National (State, Governmental) Radon Programmes, which comprehends the responsibility of the whole society for these exposures and the willingness, or the feeling of necessity, to help the citizens living with this radon risk, it means to find out these houses (by targeted searching programmes), to inform these citizens about the risk (to change the unknown risk to a known one), providing them information about certified countermeasures, and at least supporting them to realise these countermeasures (offering complete cover of expensive, some financial support, relieves, advises). Action levels have to be given, above which the adequate countermeasures are optimalized from the view of radiation protection in the country. The Radon Programme applies to only some units of percents 198 21" RHD Jasná pod Chopkom of the population, so also to a comparable part of the house stock. In Czechia these are 2%, about 200 000 citizens, or about 70 000 family houses, with an average annual exposure of 16 mSv, but with a range from 2 up to 500 mSv, with a probability of lung cancer incidence from radon 3.5 to 170 times [4] higher than the average value for the Czech population. In these houses the radon risk is caused by a combination of several factors: by the high radon risk in the soil below the house, by the imperfect (or absent) barrier between the ground and the living space and by the inconvenient ventilation regime of the users of the house, therefore at least two factors can be improved. b) Prevention of the risk For the Radon Programme an inevitable part has to be the preventive approach ensuring that henceforth houses with radon risk will not be built in the country. This could be left on the responsibility of the prospective houseowners to ensure healthy conditions in the future for the own family members or for tenants. Unfortunately the perception of risk of lung cancers from radon is very low or negligible in the population of our country, as it is in others. In that moment when the investor has to spend in addition about one percent of the expenses of the house (or to give up of some less necessary part of the house or in it), he decide as a rule unpreventively, irrationally. This can be explained in such a way that the radon risks did not come into the awareness of the population to respond adequately and knowingly. Unfortunately the radon risk is not impressed in the mind of designers and building contractors also. There is no other way for the overbridging of this irresponsibility as by a legal duty, i.e. to require this prevention by the building codes. This experience was noticed also during the absence of such a duty in our legislacy - from the invalidation of the decree No. 76/1991 by the Atomic Act in July 1997 up to the validation of this duty by the novelisation of the Atomic Act through the novelisation of the Building Act in July 1998 [5]. Really no investor requested the measuring of the radon potential of the lot, although only from this measuring he could get a quantitative measure of the radon risk needed by the designer to design an adequate preventive radon countermeasure according to the national standard ČSN 73 0601, and not to design a surplus one and therefore dearly paid for. Part of the complex preventive approach is also the preventive care over building materials and water. In this direction the Czech Atomic Act imposes important duties on producers and suppliers - to measure and to interpret the content of radium 226 in building materials and the content of radon 222 and of other natural radionuclides in water, and to announce the results to the office, including the supervision of the performance of this duties. The providing decrees set the action levels and levels above which supplying are forbidden. An important part of the prevention has to get a programme of adequate, qualified and systematic education of the population and of the specialists. The professional education can start in public schools, continue in high schools and universities with a view to civil engineering or ecology, culminating in postgraduate training courses for designers and mitigating contractors. Qualified textbooks and technical literature must base the education. Only in such a way radon will get for citizens, investors, designers and 2ľ RHD Jasná podChopkom 199 building contractors an adequately accepted and considered condition factor as dry, warm, light, silence etc feeling in a dwelling, in a house, at home. Radon Programmes for workplaces In this sphere the radiation protection is still starting and looking for the systematic approach. It is clear, that a common worker is, from the point of view of radiation protection, someone between a member of the population and a radiation worker, with radiation limits between 1 mSv and 20 mSv per year. A limit of 6 mSv is recommended for this group [6]. From this results a radon concentration of 964 Bq/m3, rounded with 1000 Bq/m3 is derived, if annual working occupancy of 2000 h (whole working load), average lung ventilation at light work of 1.2 m3/h and the conventional conversion factor of 6.23 uSv/(Bq/m3) [7] is assumed. This value is given in the IBSS and also in the providing decree of the Czech Atomic Act. This solution leads to other troubles for the implementation of the radiation protection. The authority has first to find out itself all work activities where exceeding of the limit is possible [6]. A targeted searching programme must be started, similar as that for searching dwellings with radon risk, with similar strategy - to target on areas with high radon risk in the ground according to geological radon risk maps and to target on workplaces situated in cellars and basements, in houses without cellars or with imperfect barriers against the ground. Excluded can be workplaces where the production requires higher exchange of air. From previous experience several workplaces with higher radon concentrations are known - caves, water treatment plants, spas. Other types of workplaces are plants extracting metals and processing minerals (fertiliser production), where some places exist with higher concentrations of NRS resulting in higher exposure from special operations (e.g. waste management). In contrary to dwellings where the occupation time is high (in average 80% of the year, or 7000 hours annually) the occupation time at work is markedly shorter (in average 20% of the year, or 2000 hours annually). Additionally important differences in the ventilation and heating regime during and after working time have to be expected. Therefore it is impossible to get a good estimation of the exposure by simple and cheap integrating measuring devices (over days, weeks, months, a whole year). These are mostly overestimating; it may be better for three-shift running. More appropriate are devices measuring immediately or continuously with record. With regard to the situation that the employer has not the duty for licence application he must not put before the authority results of measuring about the fulfilment or not fulfilment of radiation protection requirements. In contacts with employers therefore one has to await rational or irrational attitudes, as are the attitudes of houseowners toward the radon risk. Sometimes one can recognise interest on the new factor of risk or apprehension about professional lung cancer and therefore with a willingness for co- operation. In other cases one have to await a ban on entry Conclusion To implement radiation protection against natural radiation sources on workplaces with unintentional and unconscious exposure to these sources of radiation is a new, but interesting and important task which need new approaches, time, staff and effort 200 21" RHD Jasná pod Chopkom References [1] International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, IAEA Safety Series No. 115-1,1994 [2] Act No. 18/1997 Coll. on PeaceM Utilisation of Nuclear Energy and Ionising Radiation (the Atomic Act)..., Coll. of Law, 1997 [3] Decree of the SÚJB No. 184/1997 Coll. on Radiation Protection Requirements, Coll. of Law, 1997 [4] Health effects of exposure to radon, NRC, BEIR VI, Nat. Acad. Press 1998 [5] Act No. 83/1998 Coll. changing and amending the act. No. 50/1976 Coll. on regional planning and building codes ..., Coll. of Law, 1998 [6] Recommendations for the implementation of the European Basic Safety Standards Directive concerning significant increase in exposure due to natural radiation sources, European Communities, 1997 [7] Protection against Radon-222 at Home and at Work, ICRP Publication No. 65, Pergamon Press, 1993 2ľRHD Jasná pod Chopkom SK98K0377 201 METROLOGY OF RADON AND THORON CONCENTRATIONS Matej Ďurčik and Magdaléna Vičanová Institute of Preventive and Clinical Medicine Limbová 14, 833 01 Bratislava, Slovak Republic Introduction Many methods have been developed for the measurements of radon (222Rn) and thoron (220Rn) concentrations in air. One of the current major problems of these methods could be separation of radon from thoron. For determination of the activity concentrations are used their properties as half-life, radiation and energy of emitted particles. All measuring methods are divided into next groups: • alpha spectrometry with using semiconductor detector [1] or scintillation cell • time analysis by delayed coincidences with taking advantage of the relatively short time delay between the alpha particle emitted by thoron and its first decay product [2] • solid state nuclear track detectors with discrimination of thoron by diffusion barrier [3]. The alpha spectrometry measurements of radon and thoron concentrations by ionisation chamber, used only in laboratory conditions are described in this paper. For the measurements of radon and thoron in dwellings and work areas there was proposed diffusion double chamber detector with track detector. Materials and methods The standard system with ionisation chamber is used for laboratory measurements of radon and thoron concentrations and for calibration of radon and thoron meters and measuring techniques. The system consists of three basic parts [4]: - filling part consists of radon and thoron generators, radon tank with volume of 40 L and working gas supply (argon or nitrogen), working pressure ranges from 500 to 2000 hPa; - measuring part is cylindrical ionisation chamber with volume of 4 L. The ionisation chamber works in stationary and in flow modes. The resolution for radon alpha particles (5.5 MeV) is 180 keV and efficiencies for stationary and flow modes are in Table 1; - sampling part consists of sampling station for filling external detectors and of experimental cylindrical stainless steel chamber with volume of 10 L for exposure of passive detectors. Integral measurements of radon and thoron concentrations in the air of buildings and workplaces could be performed by double chamber dosimeter (Figurel) [5]. It consists of housing for the two chambers with wearing bracket, a track detector (CR-39) on each chamber bottom held by a ring and the chamber covers with openings for air exchange. Paper difíusion barrier (thickness 0.15 mm) was used for detection of radon and thoron concentrations and polyethylene filters (thickness 0.01 and 0.05) were used for detection of radon concentration. Chemical etching process (solution 25 % NaOH, temperature 70 °C and time 18 hours) was used for evaluating detectors. 202 2 ľ RHD Jasná pod Chopkom cover with openings diffusion barriers detector holder ring alpha track detector housing wearing bracket 0 10 20 30 mm Fig.l The double chamber dosimeter and its components Results The ionisation chamber is calibrated directly using 226Ra source or using radon store tank [4]. The efficiencies for thoron measurements are calculated from efficiencies for radon and using properties of the thoron progenies. The Table 1 shows difference between radon and thoron calibration coefficients. The efficiencies of the ionisation chamber for the radon daughters in flow mode are less of 4 % as for stationary mode. The Figures 2 and 3 show the spectrum of mixed radon and thoron sample and theoretical analysis and fitting of the measured data by commercial software. Tab.l Calibration coefficient of ionisation chamber for radon and thoron measurements Calibration coefficients [%] Nuclide Radon Thoron Stationary Flow Flow Rn 98,1 98,1 98,1 First daughter 45,5 43,7 53,9 Next daughter 45,5 43,7 43,7 Total 63,1 61,7 76,0* * It is calculated only for thoron and its first daughter product The density of the tracks on detector placed in diffusion chamber depends on radon and thoron diffusion into the chamber. The radon or thoron concentration inside of the diffusion chamber is given by equation (1) 21" RHD Jasná pod Chopkom 203 where cosh L_X.R..S (|) V " sinh(i) and h is thickness of diffusion barrier, R is diffusion length and V is volume of the chamber. The growing of the radon and thoron concentrations for different barriers is shown in Figure 4. For ellimination of the thoron in chamber it is possible to use polyethylene barrier with thickness of 0,01 mm. The discriminatory efficiencies of thoron by PE filters with thikness of 0,01 mm and of 0,05 mm are 99,5 % and 99,96 %, respectively. 100 200 300 400 500 Channel Fig.2 Alpha spectrum of radon thoron and them progenies measured by ionisation chamber 2500 170 190 210 230 250 270 290 310 330 350 Channel Fig. 3 Theoretical fit of the radon and thoron alpha spectra from ionisation chamber 204 21st RHD Jasná pod Chopkom 1UU - 80 - í Paper filter 60 - a I Thoron o 7s 40 - 20 - PE filters - 0 - , 1 T—i , T - '- TTTJ—i i i inn)—r-T mni|—r~r 2 4 6 10 100 1000 Time [min] Time [min] Fig.4 Radon and thoron concentration in the diffusion chamber for different barriers Conclusion The described dosimeter is very useful for routine measurement and would be applicated in measuring of radon and thoron concentrations in caves and dwellings. Big disadvantage of the dosimeter is small holes in cover and it could not be used in dusty areas. From previous measurements of the equilibrium equivalent thoron concentrations by semiconductor detector the measured values ranged from 0,1 to 5,6 Bq.m"3 in the Slovak kindergartens were obtained. References [1] J. Bigu: Appl. Radiat. Isot. Vol. 37, No. 7, pp. 567-573,1986 [2] F. Bochicchio, G. Campos Venuti, C. Nuccetelli, S. Risica and F. Tancredi: Environment International Vol. 22, Suppl. l,pp. S633-S639, 1996 [3] C.S. Dudney, A.R. Hawthorne, R.G. Wallace and R.P. Reed: Health Physics Vol. 58, No. 3, pp. 297-311,1990 [4] F. Havlík, M.Ďurčík and D. Nikodémova: The Safety of Nuclear Energy, Vol. 1(39), No. 3, pp. 107-111,1993 [5] O. Sorenio and A. Guhr: Nucl. Tracks Radiat. Meas. Vol. 19, Nos. 1-4, pp. 395- 400, 1991 2ľRHD Jasná pod Chopkom 'mm m ™ "oľnôvo 205 yoUo/o NATURAL RADIOACTIVITY IN SLOVAK CONSTRUCTION MATERIALS AND THE INDOORS DOSE RATE FROM BUILDING MATERIALS KCabáneková, M. Vladár Institute of Preventive and Clinical Medicine, Limbová 14, 833 01 Bratislava Slovak Republic Introduction The radioactivity in the building materials, together with the radiation from the ground and the cosmic radiation, are the main sources of external radiation to the people. The increasing use of industrial wastes as building materials, which often contain enhanced levels of the radioactivity, has increased the interest in this source of population exposure [1]. For the evaluation of the external y-ray contribution to the exposure of population, the estimation of the 226Ra content as well as of the 232Th and 40K concentration in building materials and products, is essential. The building materials with high values of 226Ra coupled with the pronounced porosity of the final products, make them potential indoor Rn sources too. For keeping the population exposure as low as reasonably achievable, (recommended by the Slovak regulations), the radioactive content of primordial radionuclides in building materials and products have not to exceed 370 Bq.kg"1 of radium equivalent activity and 120 Bq.kg"' of ^Ra [2]. Materials and methods Samples of building materials (cement, stone, fly-ash, light concrete, slag, dross, sand, dolomite, etc.) used for construction of the residential buildings were collected, milled and screened with 2-3 cm sieve. After drying, the samples were stored in 450 cm3 sealed polyethylene containers for a 30 day period. All samples were measured in a 4 n geometry (Marinelli type samples) usually for 60.000 seconds. Measurements of 226Ra, 232Th and 40K concentrations were carried out by high resolution y-ray spectrometry. The primordial radionuclides 226Ra and 232Th were assessed through their progeny photopeaks 214Bi (609 keV), 214Pb (295 keV, 351 keV) ^Ac (338 keV, 911 keV) and Pb (238 keV). The specific activity of both nuclides has been determined as weighted average of their photopeaks. 40K was measured directly via its 1460 keV peak [3,4]. Usually the natural radioactivity of building materials is specified on a common index called "Radium Equivalent Activity" estimated as a*, = am+ 1.43 an, + 0,077aK where a«, - radium equivalent activity in building material, Bq.kg"1 aRa, an,, aK - specific activities of 226Ra, 232Th and 40K in building material, Bq.kg" 206 21" RHD Jasná pod Chopkom External dose from building materials The annual indoor effective dose resulting from building materials was calculated following the procedures described by UNSCEAR [5]. The total dose fraction, expressed in mSv.y'1 results from the addition of three terms. The first represents the dose rate resulting from irradiation by external sources and the second and third concern the dose rate from inhaled radon and thoron progeny in the indoor air [6]. The determination of the external irradiation from building materials is described by the expression: 6 D = p.T.b.10" .aiui + q-rh-ami ).nii where D - external irradiation dose rate (mSv.y"') p - fraction of time spent indoors, 0.8 T - 8760 h per year b - conversion from absorbed dose in air to effective dose equivalent, 0.7 Sv per Gy mj - mass fraction of type i material in reference room aKi aRai - specific activity of 2MRa, 232Th and 40K (Bq.kg'1) in type i building material ára and qK, qRa, qra are the conversion factors from the specific activity of the building materials to absorbed dose rate in indoor air ( nGy.h"1 per Bq.kg"1). The choice of these conversion factors is the most important factor for evaluating of the external dose from building materials. Because the gamma radiation from walls is strongly dependent on the wall thickness and on the density of building materials, a standard room must be assumed to estimate the dose rate. Table 1. shows some values of these coefficients for different model rooms. Tab. 1. Values of the conversion factors to evaluate the external dose for different room models Wall Specific dose rate Reference Room geometry density rnGv.K1DerBa.kě1l Wall thickness ľK-cm*] X Stranden8 9x5x2.5 n? 2.32 0.88 1.04 0.08 20 cm Stranden* 4x5x2.8 n? 2.35 1.05 1.27 0.09 20 cm Koblingef* 4x5x2.8 n? 2.35 1.05 1.18 0.09 20 cm Koblinger 4x5x2.8 n? 2.35 0.7 0.84 0.06 20 cm Koblinge? 4x5x2.8 n? 1 0.59 0.81 0.06 20 cm Ackers7 6x4x3 m3 2.32 0.62 0.89 0.05 20 cm 21" RHD Jasná pod Chopkom 207 Results and discussion Until now, in our Institute about 600 samples of building materials have been measured. The measured activity concentrations and calculated radium equivalent activities of the materials, which are one of the most used in the Slovak building trade are given in the Table 2. The average values of natural radioactivity for some Slovak building materials (cement, fly-ash, concrete, brick) are approximately on the same level as in other countries [10,11,12]. The data in Table 2. show broad range of concentrations of the natural radionuclides. This fact reflects their natural origin and geological conditions at the site of their production. The construction materials with concentrations higher than recommended in Slovak regulations, are forbidden for using in the building trade. Tab. 2. The specific activities of li6Ri, !!Th, "k and radium equivalent activity in various types of building materials in Slovak Republic A,Bq/kg Material N 2siTh nc nun. AVO max. min. AVG max. min. AVG max. min. AVG max. Cement 88 16 42 116 12 21 37 161 249 352 46 91 196 Stone 96 1 16 106 1 16 133 2 342 2441 2 65 484 Fly ash 92 34 102 318 21 70 159 33 458 937 67 237 618 Slag 17 44 86 1S8 39 64 106 149 287 575 111 200 354 Dio& 5 75 163 257 14 37 51 76 295 650 101 238 381 Sand 14 4 28 258 1 21 142 5 339 912 5 84 .531 Concrete 29 10 60 127 5 36 70 96 336 652 25 138 278 Brick 25 12 71 265 13 80 252 72 590 887 36 231 694 MarbJe 7 1 14 41 0.3 37 95 4 504 1323 2 105 278 For a typical Slovak home made basically of brick, brick(slag) and concrete with the mean concentrations of 226Ra, 232Th and 40K given in the Table 2., the mean annually effective doses are shown in Table 3. The external dose rates were calculated by Acker's conversion factors and the values are in the range between 0.58 and 0.69 mSv.y'1. These values are a little higher as the data from the Spanish houses [6]. Tab. 3. The values of the mean annual effective doses of gamma radiation in various type of buildings Eff.Dose Material [%] Type of building imSv/yl Brick Brick Concrete Sand Total (SJBS) Brick 0.70 0.00 0.15 0.15 0.58 Brick (Slag) 0.00 0.70 0.15 0.15 0.52 Concrete 0.95 0.00 0.00 0.05 0.69 The philosophy of radiation protection suppose that radiation doses from all sources should be kept as low as achievable. Our control mechanism can regulate that part of the radiation exposure to the public which is due to the natural radioactivity present in building materials. From the measured levels of 226Ra and from the knowledge of the contribution of the external terrestrial radiation, of the building design, of the ventilation rate and of the exhalation rate of the final products, the equilibrium equivalent concentration of 222Rn can be also assessed. 208 21" RHD Jasná pod Chopkom References 1. ICRP. Principles for Limiting Exposure of the Public to Natural Sources of Radiation. Publication No.39. (Oxford: Pergamon), (1983) 2. V.K.Shuka, S.Sadasivan, V.K.Sundaram, S.V.Nambir: Assessment of Gamma Radiation Exposure Inside a Newly Constructed Building and a Proposed Regulatory Guideline for Exposure Control from Natural Radioactivity in Future Buildings. Rad.Prot.Dos. Vol.59, No.2,pp. 127-133, (1995) 3. A.Savidou, C.Raptis, P.Kritidis: Natural Radioactivity and Radon Exhalation from Building Materials Used in Attica Region, Greece. Rad.Prot.Dos., Vol.59, No.4., pp. 309-312,(1995) 4. H.Cabáneková: Concentration of Natural Radionuclides in Various Types of Building Materials in Slovakia. Jour.of Radioanalytical and Nuc. Chem., Vol.209, No.2, pp. 301-306,(1996) 5. UNSCEAR Report 1988 6. L.S.Quindos, G.J.Newton, M.H.Willkening: On the Dose Rate Indoors from Building Materials. Rad.Prot.Dos. Vol.19, No.2, pp. 125-128, (1987) 7. J.G.Ackers, B.F.Bosnojokovic, L.Strackee: Limitation of Radioactivity Concentrations in Building Materials Based on a Practical Calculation Model. Rad.Prot.Dos. 7, pp. 413-416 (1984) 8. E.Stranden: Radioactivity in Building Materials. NRPB, Didcot, Oxon (1982) 9. L.Koblinger: Mathematical Models of External Gamma Radiation and Congruence of Measurement. Rad.Prot.Dos. 7, pp. 227-234 (1984) 10. D.Popovic, G.Djurcic, D.Todorovic: Radionuclides in Building Materials and Radon Indoor Concentration. Rad.Prot.Dos. Vol.63, No.3, pp.223-225 (1996) 11. Ch.Schuler, RXrameri, W.Burkart: Assessment of the Indoor Rn Contribution of Swiss Building Materials.Health Physics Vol. 60, No.3, pp. 447-451, (1991) 12. Man-yin W. Tso, Choir-yi Ng, J.K.C.Leung: Radon Release from Building Materials in Hong Kong. Ill 2ľ RHD Jasná pod Chopkom SK98K0379 209 RADIATION LOAD FROM RADON EXPOSURE IN SLOVAKIA Magdaléna Vičanová, Matej Ďurčik, Denisa Nikodémova Institute of Preventive and Clinical Medicine Limbová 14, 833 01 Bratislava, Slovakia Introduction Slovak National Radon Program (begun in 1991) was organised by the Institute of Preventive and Clinical Medicine in Bratislava in co-operation with the regional Specialised State Institute of Public Health. Radon concentrations and radiation load were investigated in dwellings (family and multifamily), schools, public buildings, spa buildings, caves and mines. In 1992 the Slovak Ministry of Health published an ordinance about the protection of population to radon and other natural radioactive sources. The action level of annual average equilibrium equivalent concentration (EEC) of radon was established as 200 Bq.m'3 for existing dwellings and 100 Bq.m"3 for built in the future. The action level of radon concentrations for water sources is 50 Bq.ľ1 Materials and Methods Passive solid state nuclear track detectors (SSNTD type CR-39) were used to measure indoor radon concentrations. Detectors were placed in about: • 6,000 selected dwellings (minimum two detectors for every residence) • 1,000 selected buildings of the kindergartens and basic schools • 12 selected spa buildings and distributed by professional workers from regional Specialized State Institutes of Public Health. Questionnaires for obtaining the detailed information required for interpretation of results were sent to each building and house. After six months exposed detectors and questionnaires were returned to IPCM for analysis. Personal doses were measured by a pair of SSNTD in a passive two chamber system at: • 130 miners from three ore mines • 13 turist guides from seven show karst caves. Electrochemical etching combined with a chemical pre-etching process was used for evaluating detectors. Track counting is performed with the image analysis system QUANTIMET 520. The calibration of detectors was carried out in the reference radon and radon progeny measuring chamber, at the State Metrological Centre of IPCM. Experimental Results Our present results are from 3,657 residences (0,2% of total dwellings in Slovakia). It was found that the arithmetic mean (AM) of EEC was 86±119 Bq.m , the geometric mean (GM) was about 41±2.22 Bq.m and 11% of dwellings (N=409) have a greater EEC of radon than the action level. 210 2ľ' RHD Jasná pod Chopkom The arithmetic and geometrie mean of EEC varied between type of building and districts of Slovakia. The sample of family houses (N=2,363) has AM 125±135 Bq.m"3, GM 73±1.8Bq.m"3 and the sample of multifamily houses (N= 1,294) has AM 22±24Bq.m"3, GM 15± 1.46 Bq.m"3. We calculated the population-weighted AM of EEC for every district by different type of house, and then estimated this value for the whole of Slovakia obtaining a figure of 48 Bq.m"3. The annual average effective dose is 2.1 mSv per inhabitant and the proportion of indoor radon exposure in each district of Slovakia is presented in Figure 1 and distribution of radon levels in Table 1. ä 1.4-2.S 12.8-4.2 1 4.2 - 5.6 Fig.l Annual average effective doses from indoor radon exposure in districts of SR Tab.l. Distribution of indoor radon concentrations in Slovakia EEC Number of Remedial [Bq.m-3] dwellings [%] Actions <200 88.6 200 - 599 10.6 to 10 years >600 0.8 to 3 years The maximum value of EEC which was found isl500 Bq.m"3 The radon survey in the school was performed in 645 buildings (10.8 % of all schools and kindergartens). It was found, that the action level of EEC of radon was exceeded in 16 schools (13 kindergartens and 3 basic schools). The annual effective doses from radon exposure (E) were estimated for children, students and teachers (Tab3). Personal doses (miners) were measured monthly during 1995 y and personal doses (tourist guides) were investigated during three years period (1995-1997) (Tab.3). The radiation load from radon exposure of spa staff (Tab.3) was estimated from investigation of radon concentrations in the air of spa buildings (Tab.2). 21st RHD Jasná pod Chopkom 211 Tab.2. Investigation of Radon Concentrations in the Air of Spa Buildings Spa Place EEC [Bq.m-31 Bojnice Baník 30 ±8 Mier 439 ±162 Brašno bathroom 106 ± 53 Číž bathroom 101 ± 24 spring 464 ±125 Dudince Rubín 64 ±9 Korytnica bathroom 96 ±31 Kováčova bathing-pool 395 ±103 relaxation-room 142 ± 34 Lúčky bathroom 980 ± 274 Nimnica bathroom 32 ±9 Rajecké Teplice bathing-pool 90 ±44 Sliač bathroom 232 ± 58 Sklené Teplice Public room 1180 ±330 Central hall 1110 ± 311 Parenica 838 ±226 Turčianske Teplice Blue bathing-pool 264 ±145 Tab. 3. Annual Average Effective Doses (E) from Radon Exposure Place ^Rn Time" [Bq.nť3l [hi fmSvl Outdoor 15 1760 0.011 Indoor-SR 120 7000 2.1 Indoor - world 40 7000 0.7 Kindergarten 144 teachers 1500 0.52 children 1000 0.35 Schools 112 teachers 1500 0.40 students 800 0.22 Mines 710 2000 4.5 Caves 7680 staff 400 9.6 visitors 1 0.024 spa - staff 965 2000 6.11 1} Assuming time 2) Conversion factors (for houses and workplaces) from BSS 1996 equilibrium factor 0.4 212 2ľ'RHD Jasná pod Chopkom Conclusions The national survey results suggest that SR may be among the countries with higher radon risk in Central Europe. The annual average effective dose from indoor radon exposure is 2.1 mSv per inhabitans. The districts with highest indoor radon concentrations and districts with high radon levels in spa buildings correlate with known presence of uranium in the soil. The soil is probably the main source of radon in Slovak dwellings, spa and scholl buildings too. Our next aims are to continue with measurements of radon levels in dwellings and spa areas too, and to investigate the radiation load in spa staff resulting from radon exposure The results from investigation of personal doses underground workers suggest that the Slovak caves should be designated as a controlled area according to BBS 96. Acknowledgements We are grateful to all professional workers from regional Specialized State Institutes of Public Health for their help at distribution radon detectors, we thank Mrs.Oľga Mlynárova and Mrs.Larisa Kulková for their excellent technical help. References /. Vičanová M., Ďurčík M., Nikodémova D.:"Indoor Radon Exposure of Slovak Population." European Conference Protection against Radon at Home and at Work, Book of Proceedings, Prague,June 1997 2. Vičanová M.,Nikodemová D., Ďurčík M.,Havlík F.: Indoor Radon Concentrations in "Hot Spots" locations in Slovakia. Proceedings of Healthy Buildings '95, Miláno 10-14 September 1995, Vol.2, po.715-720 3. F.Havlík, M.Ďurčflk and D.Nikodemová, State Metrological Centre of Slovakia for Radon Quantities, Bezpečnost jadrovej energie, 1(39), 1994, No.4. 4. Gombala E., Ďurec F., Blazseková, Vičanová M.: Radón v ovzduší kúpeľov stredoslovenského regiónu. Rádioaktivita v životnom prostredí. Spišská Nová Ves 1997 5. UNSCEAR-Report 1993, Sources and Effects of Ionizing Radiation, United Nations, New York 1993. 6. Inter. Basic Safety Standards for Protection against Ionising Radiation, Saf. Ser. 115, IAEA Vienna, 1996 7. Sarenio O., Guhr A.,: „A passive individual dosimeter for integrating measurements of the radon daughter product contents in air", Nucl.Tracks Meas., Vol.19, No 1-4, pp. 395-400,1991 8. Vičanová M., Nikodémova D., Ďurčík M., Havlík F.: Radon Air concentrations in karst caves and estimated of personal doses at employess. IRPA regional symposium. Radiation protection in neighbouring countries of Central Europe. Prague 1997, pp. 195-198 9. Ďurčík M., Havlík F., Vičanová M., Nikodémova D.: Radon Risk Assessment in Slovak Kindergartens and Basic Schools. Rad.Prot.Dosim., Vol.71, No.3., pp 201- 206 (1997) 2ľ'RHDJasnápodChopkom SK98K0380 213 2MRn CONCENTRATION IN THE OUTDOOR ATMOSPHERE AND ITS RELATION TO THE ATMOSPHERIC STABILITY K Holý1}, R. BôhmI}, I. BosáI}, A. Poláškovál), O. Holá 2) '* Faculty of Mathematics and Physics ofComenius University, Mlynská dolina Fl, 84215 Bratislava, Slovak Republic, 2> Faculty of Chemical Technology of Slovak Technical University, Radlinského 9, 8123 7 Bratislava, Slovak Republic Introduction The radon (222Rn) concentration in the outdoor atmosphere is not stable. It was found out in many studies that the 222Rn concentration present the daily and seasonal variations in the air near to the earth's surface [1]. These variations are ascribed to variations in atmospheric dispersion conditions [2]. Therefore, in the last period the important motivation for the study of radon in the outdoor atmosphere is also the possibility of the utilize of 222Rn in atmospheric studies, especially for the determination of the atmospheric stability [3-5]. In the meteorology, the atmospheric stability indexes are joined to the set of the meteorological parameters which influence on atmospheric dispersion conditions. In the present time, especially, the following methods of the determination of the atmospheric stability are used: Pasquill's method, Turner's method and their modifications [6,7]. These methods give also only an approximate information about the atmospheric stability. The direct continuous measurement of the atmospheric stability, for example the measurement of the temperature gradient by the usage of the high mast, is rather difficult and expensive. The atmospheric stability classification originates mainly from the following meteorological parameters: the solar radiation, the wind velocity and the cloudiness. These are the same parameters which influence on the radon concentration in the outdoor atmosphere [2,8]. In the last period we have tried to find the relation between the radon concentration in the surface air and the vertical atmospheric stability of the lower atmosphere. The preliminary results are very optimistic and they are presented also in this contribution together with some results of the investigation of the radon variations. Methods The measurements were carried out on the open grass area in campus of Faculty of Mathematics and Physics in Bratislava at a height of 1.5 m above the ground surface. The radon in the outdoor atmosphere has been monitored continuously using the large volume (~ 4.5 I) scintillation chamber [9]. The radon-sampling device consists of the air transport tube (~ 30 m), the water trap (- 20° C), the flow-rate meter, the aerosol filter, the large-volume scintillation chamber, the gas meter and the pump. The flow-rate of air through the radon-sampling device was 0.5 // min and it was selected in such a way that 22Tln was decayed still before the inlet of air into the scintillation chamber. The record of detector's data has been made every 30 minutes. The 222Rn concentrations have been calculated by means of the method published by D. C Ward and T. B. Borak [10], based on the data corresponding to the time interval of 2 hours. 214 2 ľ RHD Jasná pod Chopkom Results and discussion The radon in the outdoor atmosphere has been monitored continuously since 1991. On the basis of the measured data mainly the average daily and the average annual courses of the 222Rn concentrations have been studied. Hour of day 0-24 0-24 0-24 0-24 0-24 0-24 11- -J ' ' i i i i i i i i i 0-24 0-24 0-24 0-24 0-34 0-24 10- 9- • .- 8 \ : . .. • • t \ • c 6 o * ^ 1 5' J i: Jan Feb Mar Apr May Jun Jul Aug &p Oct Nov Dec Month Fig.l. The average annual course of the 222Rn concentration for years 1991-1997 (solid line) and the average daily courses of the 222Rn concentration for the individual months (dashed line). The annual courses of 222Rn concentration are similar for all years. They present the annual variations. The average annual course of the 222Rn concentration calculated on the basis of all continual measurements in years 1991 - 1997 reaches the maximum value in October and the minimum value in April (Fig.l). In this figure, there are shown also the average daily courses of the 222Rn concentration for the individual months of the year. These average daily courses have a form of waves with a maximum in the morning hours and with a minimum in the afternoon. The maximal amplitudes of daily waves have been reached in the summer months, from June till August. The amplitudes of daily waves are very small at the end of an autumn and during the winter months. The analysis of the daily waves and annual courses of 222Rn showed that the amplitudes of the daily waves are in proportion to the global solar radiation irradiating the Earth's surface. The day duration influence on the phase of the daily wave and the wind velocity influence mainly on the level of the radon concentration [8,11]. For the study of the relation of the radon concentration in the outdoor atmosphere to the atmospheric stability the data of the atmospheric stability had to be obtained. However, the atmospheric stability is not directly measured in the vicinity of our faculty. Because of this reason we calculated the atmospheric stability on the basis 21" RHD Jasná pod Chopkom 215 of the meteorological data obtained from Department of Meteorology and Climatology [12] according to Turner's method modified by Nesler and Reuter [13]. This classification distinguishes seven stability classes: 1 - extremly unstable, 2 - unstable, 3 - slightly unstable, 4 - neutral, 5 - slightly stable, 6 - stable, 7 - extremly stable. These stability classes were determined on the basis of the wind velocity and irradiation indexes where the last ones were calculated by using of the data of the sun height and the cloudiness. The night cloudiness we obtained by the extrapolation of the meteorological observations at 9 p.m. and 7 a.m. and the daily cloudiness we obtained from measured data of the sunshine duration [14]. 10 I.I.I I . I . I . I . I r7 9- o -^o- o o o -6 s r-, 8- / / -5 7 / i / • . • • \ V 6- 9-.- " 5- a 4- o o -o— o 2% -1 A A A- -0 -1 0 2 4 6 8 10 12 14 16 18 20 22 24 Time [h] Fig. 2. The variation of the 222Rn volume activity, the wind velocity and stability index in the outdoor atmosphere in August 1997 (»- the 222Rn volume activity, o - the stability index, A - the wind velocity, —, -.- the results of the Fourier analysis) In Fig.2 there are compared the average daily course of the radon concentration and the average daily course of the atmospheric stability in August 1997. Also theaverage daily course of the wind velocity is shown in this figure. August 1997 was a little cloudy and the probability of the incorrect determination of the stability was very low. We can see that the courses of the radon concentration as well as the atmospheric stability are approximately the same. The minimum of the vertical atmospheric stability and the minimum of the radon concentration are reached in early afternoon hours. The both parameters reach their own maximum values at the night hours. There is a shift between the atmospheric stability and the radon concentration. This shift could be approximately 2 hours but this effect was not analysed in detail. The correlation between both courses is high (the correlation coefficient R = 0.98) although the radon data were not shifted.. The analogical high correlation between the radon concentration and the atmospheric stability was found out also for other months of year 1997 [15]. The results indicate that the 222Rn concentrations in the outdoor atmosphere could be used for determination of the vertical atmospheric stability and these ones 216 21st RHD Jasná pod Chopkom could reflect the atmospheric stability more completely than the different classifications based on the knowledge pertinent to the meteorological parameters. Acknowledgements. This study was funded by Scientific Grant Agency of Ministry of Education of Slovak Republic ( VEGA project No. 1/4194/97 ) and International Atomic Energy Agency, Vienna, Austria (Res. Contract No: 9093/RO). References fl] Gessel, T.F.: Health Physics, Vol. 45, No. 2,289-302 (1983) [2] Garzon, L., Juanco, J.M., Perez, J.M., Fernandez, J.M. and Arganza, B.: Health Physics, Vol. 51, No. 2, 185-195 (1986) [3] Dueňas, C, Pérez, M., Fernandez, M.C., Carretero, J.: J. Geophys.Res., Vol. 99, No. D6,12865-12872(1994) [4] Fujinami, N., Esaka, S.: J. Geophys.Res., Vol. 92, No. Dl, 1041-1043 (1987) [5] Dueňas, C, Pérez, M., Fernandez, M.C., Carretero, J.: J. Environ. Radioactivity, Vol. 31, No. 1,87-102(1996) [6] Pasquill, F.: Atmospheric Diffusion, Van Nostrand Reinhold, New York, 1962 [7] Turner, D. B.: J. Appl. Meteorol, 3, 83, (1964) [8] Holý, K., Bôhm, R., Polášková, A., Holá, O.: Proč. of Int. Congress on Radiation Protection IRPA 9, Vienna, Austria 1996, Vol. 2,176-178 (1996) [9] Beláň, T., Chudý, M., Ďurana, L., Grguľa, M., Holý, K., Levaiová, D., Povinec, P., Richtáriková, M., Sivo, A.: Rare Nuclear Processes, Proc. of 14th Europhysics Conf. on Nucl. Phys., World Scientific Publishing, Singapore, 345- 366 (1992) [10] Ward, D. C, Borak, T. B.: Health Physics, Vol. 61, No. 6,799-807 (1991) [11] Holý, K., Bôhm, R., Polášková, A., Holá, O.: Long-term measurement ofRn-222 concentrations in outdoor atmosphere (poster), The IRPA Reg. Symp. on Rad. Prot., Sept. 8-12,1997, Prague, Czech Republic [12] Tomlain, J., Hrvol, J.: Ročenka meteorologických meraní a pozorovaní za rok 1997, Meteorologické observatórium KMK, MFF UK, Bratislava, 1998 [13] Gajar, B.: Comparison of different atmospheric stability classification, Dep. of Met. Com. University, Bratislava, 1974 [14] Hrvol, X: Acta Met. Univ. Comemianae, Vol. XXIII, 12-23 (1994) [15] Holý, K., Chudý, M., Sivo, A., Polášková, A., Richtáriková, M., Bôhm, R., Petruf, P., Holá, O.: I4C and 222Rn concentrations in the outdoor atmosphere and in the soil air, FRCM of the IAEA CRP on „Isotope - Aided Studies of Atmosperic Carbon Dioxide and other Greenhouse Gases", Groningen, Netherlands, Sept. 8-11,1998 2ľ RHD Jasná podChopkom SK98K0381 217 CONTINUAL MONITORING OF RADON DECAY PRODUCTS CONCENTRATIONS IN INDOOR AND OUTDOOR AIR P. Petruf, K. Holý, T. Stany s Department of Nuclear Physics, Faculty of Mathematics and Physics, Comenius University, Mlynská dolina Fl, 84215 Bratislava, Slovakia 1. Introduction The extensive research of behaviour of radon progenies in the indoor and outdoor air has been made in last years. A precise and quick method for the determination of radon daughters in the air is needed to solve some problems in health hazard estimation in the indoor air [1,2] and in the research of the atmospheric stability or transport of air masses in meteorology [2,3]. The goal of our work was the development of the method and construction and testing of measurement device for continual monitoring of radon daughters concentrations in the indoor and outdoor environment with regard to make possible to determine very low volume activities in the outdoor air (below 5 Bq/m3). Some results have been already presented earlier [4]. 2. Materials and methods For daily observation of the radon decay products concentrations the monitoring equipment must be simple to operate and capable to work in the considerable background radiation field. For this purpose the measurement of an alpha activity of material deposited on the filter surface is most suitable [4]. We have chosen a three- count filter method (sometimes called modified Tsivoglou method [5]) for the determination of the activity concentrations of the radon decay products in our experiment. In this method air sample is drawn through the appropriate filter material. Radon and thoron daughters both attached and unattached on aerosols particles are collected on the filter surface and then the filter activity is counted. We have used silicon surface barrier detector with the active area of 200mm2 in our monitor. We have chosen a Millipore1 AW19-type filter and sampling rate of 30 1/min for collecting of the air samples. The choices have been based on results of our earlier experiment [4]. The whole sampling and measurement procedures are controlled by a personal computer, so the monitor can work automatically. The basic scheme of our monitoring equipment is in the Figure 1. The determination of the individual activity concentrations in three-count method is based on the solution of the simultaneous equations describing the number of atoms of measured nuclides on the filter during and after sampling. The original three-count method has a several modifications [4]. Theoretical considerations and experimental results indicate that the higher statistical precision can be obtained by using of the spectrally resolved alpha-counting method [6]. The increase of the precision of the 2l8Po (Ti/2=3.05min) volume activity determination can be achieved by utilising of the 1 Millipore Corp., 397 Williams St, Marlboro, USA. 218 2 ŕ RHD Jasná pod Chopkom sampling time interval as the first counting one [5, 7]. The system of equations for the computing of the individual radon progeny volume activities by the spectrometric method may be described in a matrix form as: |M|> (1) els2vf Figure 1 - The scheme of the monitoring equipment for the determination of radon decay products in the air. The location of the detector near the filter makes possible to count the alpha particles emitted from the filter surface during a sampling. where A is a 1x3 matrix, which coefficients are activity concentrations of radon progenies, \M\ is a 3x3 matrix which elements are functions of times of starts and ends of the sampling and counting intervals, Ei, E2 and S3 are the counting efficiencies for alpha-particles of 6 MeV (2l8Po), 7.68 MeV (2I4Po) and 8.78 MeV (2'2Po from thoron decay chart), fis the filtering efficiency, v is the sampling rate and Ni, N2 and N3 are the numbers of counts obtained in the individual counting intervals [4]. Considering the Poisson distribution of numbers of counts and neglecting the correlation between numbers of counts in the individual time intervals we can compute the standard uncertainty of activity concentrations as: a\A,) = v/- | (2) where ij= 1,2,3 respectively, for individual radon daughters, Aj are coefficients of vector A, my are coefficients of matrix M and a is symbol for standard uncertainty. The shape of the equations (1,2) is the same also for the method not using the sampling 2 ŕ RHD Jasná pod Chopkom 219 time interval for counting. Differences are only in the shapes of the coefficients of the matrix M. More details of the mathematical computing technique of radon decay product volume activities can be found in several publications [4-6]. The following time intervals have been chosen for the determination of the radon daughters activity concentrations: the time of 20 min for sampling and counting of the 218Po and 214Po alpha particles (Nj, N2) and after a pause of 15 min then follows the 40 min time interval for the additional counting of alpha particles from 2MPo (N3). The number of counts of alpha particles, which have an origin in the thoron decay chart, was subtracted from the number of counts in the 6 MeV peak using the 8,78 MeV alpha- particles counting [4]. Apr 23 Apr 24 Apr 25 Apr 26 Apr 27 date Figure 2 -The daily variations of the activity concentration, air temperature and pressure. The monitor was tested in three different environments: • in the basement of the building, • in the room on the 2nd floor of the same building, • in the outdoor air in front of the building. The samples were taken every two hours in case of basement and outdoor air measurements and every four hours in the measurements on the second floor. Simultaneously with the radon progenies, the radon activity concentrations were measured. The ALPHA GUARD PQ2000/MC50 monitor and the large volume scintillation chamber [8] were used in the indoor air and in the outdoor air measurements, respectively. 220 21" RHD Jasná pod Chopkom 3. Results and discussion The average values of measured activity concentrations and their standard uncertainties are in the Table 1. Our results show an agreement with the expectations of the higher activity concentrations in the indoor air in comparison with the outdoor air and also the decrease of the concentrations in the higher floors compared with the basement. The relatively low standard errors of mean activity concentrations in the basement are obviously related to the high stability of the air conditions and it also could be expected. In the indoor environments the activity concentration depends mainly on an intensity of the air masses exchange between the indoor and outdoor environment. On the second floor, where the ventilation is relatively high, the relative standard errors were on the outdoor air levels. Naturally, the higher standard uncertainties of the individual values measured in the outdoor air must be taken into account. Environment Basement 2nd floor Outdoor air (60 samples) (25 samples) (70 samples) Activity con. SU Activity con. SU Activity con. SU 3 3 3 (Bg/m ) (%) (Bq/m ) (%) (Bq/m ) (%) ^Rn 61,4 ±5,0 9,6 22,2 ± 7,9 17,6 4,1 ±2,7 37,6 29,5 ±2,8 5,4 7,3 ±2,8 9,6 2,3 ±0,9 17,4 "4Pb 14,1 ±1,8 10,6 4,6 ±1,9 13,0 1,5 ±0,8 20,0 '"Bi 12,1 ±1,6 8,7 2,6 ±1,2 15,4 1,4 ±0,6 21,4 Table 1 - The average values of the activity concentrations of radon and its decay products in three different environments. The average values are accompanied by their standard errors. In SU columns there are average values of individual standard uncertainties. The daily variations of the activity concentrations of radon and its decay products and the air temperature and air pressure in the outdoor air are illustrated in Figure 2. For first two days we can clearly see the daily variations of the activity concentration with a maximum in early morning hours and a minimum at the midday. These days were sunny and the air pressure was relatively high. That indicates the lower stability of the atmosphere in the boundary layer near the ground level and the higher intensity of the vertical air mixing. The stability of the atmosphere in last days of our experiment was higher, which causes only weak variations in the activity concentration. Conclusions The monitor of radon decay products was tested. The results show a good agreement with our expectations of the activity concentrations in three different environments. The monitor enables to determine low activity concentrations in the outdoor atmosphere with an acceptable precision during one hour counting. So, the monitor can be used for the research of the correlation between the atmospheric stability and activity concentrations of radon decay products. The same monitor and similar method (with differences only in number and shape of the matrix coefficients in formulas 1 and 2) can be used for the thoron progenies determination. However, this measurement is more time consuming because of the relatively high decay half-time of 212Pb. 2ŕRHDJasnápodChopkom 221 Acknowledgements This study was funded by Scientific Grant Agency of Ministry of Education of Slovak Republic, IAEA Vienna, Swiss Fund for Science and Scientific Fund of Comenius University. References 1. J. Porstendôrfer: Protect. Dosim. 7:107-113 (1984). 2. J. Porstendôrfer, G. Butterweck, A. Reineking: Health Physics, 67(3): 283-287 (1994). 3. C. Duenas, M. Perez, M.C. Fernandez, J. Carretero: Journal of Geophys. Research. 99(D6): 12865-1272 (1994). 4. T. Stanys, K. Holy, P. Petruf: Acta Physica Universitatis Comenianae, (1998), in press. 5. Ch. Zhang, D. Luo: Nuclear Instruments and Methods, 215:481-488 (1988). 6. N. P. Thiessen: Health Physics, 67(6): 632-640 (1994). 7. P. Vojtyla, M. Chudý, K. Holý, A. Šivo, A. Polášková, M. Richtáriková, R. Bohm, P. Petruf, M. Futas: Progress report for IAEA research contract 909B/RA; Comenius University, Faculty Mathematics and Physics, Department of Nuclear Physics, Bratislava, Slovakia (1997). 8. K. Holý, R. Bôhm. A. Polášková, O. Holá: Proč. of Int. Congress on Radiation Protection IRPA9, Vienna, Austria 1996, Vol.2,176-178 (1996). 222 SK98K0382 2ľ RHD Jasná pod Chopkom RADON IN WORKPLACES - APPLICATION OF NEW SLOVAK LEGISLATION Marek Futas', Eduard Gombala 2 1 Štátny zdravotný ústav SR, Trnavská 52, 826 45 Bratislava 2 Štátny zdravotný ústav, Cesta k nemocnici 1, 975 56 Banská Bystrica In 1996 Slovak parliament adopted changes in the Health Protection Act No. 272/1994. This amendment which came into force on January 1st 1997 deals mainly with the basic principles of health protection against ionizing radiation. It is fully compatible with international radiation protection standards [1,2] and introduces ALARA principle, justification of practice, optimization of protection and individual dose limits of 20 and 1 mSv.year'1. According to provisions of this act [3] "Workplace with ionizing radiation sources is ... workplace where radon concentration exceeds 1000 Bq.m"3". This fundamental provision gives a tool to cope with the professional radon exposure. The chosen action level 1000 Bq.m"3 corresponds to the annual effective dose 6 mSv providing 2000 working hours spent in such workplace [4]. Radon concentration and personal dose measurements conducted by Vičanová et al. [5] in 1995, 1996 proved that aforementioned action level may be exceeded in underground workplaces e.g. mines and caves. To carry out the categorization of underground workplaces the workgroup comprising workers of State Health Institute of the SR in Bratislava and State Health Institute in Banská Bystrica was created. During 1997 this workgroup took the set of radon concentration measurements in four ore mines and twelve show caves (see Tables 1 and 2). Table 1 Slovak ore mines Ore mine Mined ore Number of miners Hodruša - Hámre Gold 140 Talcum - Magnezit Hnúšťa Talc, magnesite 50 Želba Rudňany Barytes, polymetallic ores 100 Želba-SideritN.Slaná Siderit 480 Table 2 Slovak show caves Cave Tourist path Number of guides length (m) (professionals / voluntary workers) Belianska 1135 5/7 Bystrianska 750 3/1-2 Demänovská j. Slobody 1400/2500 5/14 Demänovská ľadová 450 3/4 Dobšinská ľadová 500 2/2-10 Domica 1775 3/4 Driny 380 2/3 Gombasecká 300 2/3 Harmanecká 720 2/2-4 Jasovská 720 2/3 Ochtinská aragonitová 250 3/4-11 Važecká 230 3/3-4 2ľ' RHD Jasná pod Chopkom 223 The grab sampling method (1000 ml Lucas cells) was combined with parallel continuous monitoring (Silená 5 S radon gas monitor) in month intervals since February till August 1997. These measurements were done in any stope, drift heading, travelway, shop, lunchroom, or any other underground location where miners work, travel, or congregate and along the tourist routes in caves. Some results are in Tables 3 and 4. Table 3 Radon activity concentration measurements in Talcum-Magnezit Hnúšfa Radon activity concentration [Bq.ni3! Location February March* April May Výklopník 108 ±9 87 51 276 M61/2/3mo 207 - 249 ±12 898 ±23 17CH/60/3mo 178 897 790 ±21 1676 ±31 19CH/59/3mo 200 - 2322 2944 ±40 Sklad trhavín 117 387 ±36 349 ±14 347 12CH/60/2mo - 1104 359 ±16 - M60/2/2mo 39 24 ±4 60 - * Mine ventilation system switched off Table 4 Radon activity concentration measurement in Važecká cave, 27 June 1997 No. of Location Radon activity concentration measurement [Bq.m3] 1 Zrútený dóm 3565 ±44 2 Jazierková dvorana 4312 ±49 3 Húskova hala 19 917 ±106 4 Galéria 24 470 ±118 According to averaged radon levels the following conclusions were made for show caves: All show caves except two ice caves shall be declared as workplaces with ionizing radiation sources for radon levels high above 1000 Bq.m"3 were detected and no technical mitigation is possible because of preservation of cave decoration. Situation in ore mines is a bit different. Measured radon levels were found varying in a wide range depending on the position in the mine, work practices and season. The increase of radon gas concentration in summer months was observed as it was expected. In every ore mine workplaces with radon concentration high above the action level were found. Unlike show caves in mines the technical measures should be the first step to reduce the exposure to radon and its progeny. Powerful ventilation system is a basic prerequisite and can reduce radon concentration below the action level. Other approaches are also available. For example high radon concentrations (up to 4200 Bq.m'3) in Hodruša Hereditary Drift (haulageway for Hodruša gold mine) were successfully reduced to 200 Bq.m'3 by walling up mouths of the old mine works. In other three mines the current ventilation conditions are not sufficient to ensure radon concentrations below the action level for the time being. This is the reason why mines in Hnúšťa, Nižná Slaná and Rudňany shall be declared as workplaces with ionizing radiation sources. This declaration is an administrative act, which enables to adopt adequate measures in radiation protection of workers. According to the statements of 224 2 ľ RHD Jasná pod Chopkom the Health Protection Act in workplaces with ionizing radiation sources is required to designate the controlled areas in which specific protective measures are required including the personal dosimetry. In each workplace the qualified expert is assigned as the supervisor for the questions of radiation protection and program for the monitoring of workplace and personal monitoring. The Institute of Preventive and Clinical Medicine in Bratislava provides personal dosimetry based on SSTD in three month monitoring intervals. In 1998 radon concentration measurements continue in underground workplaces which have not been classified yet. These are underground water power plant in Kremnica, open-air mining museum in Banská Štiavnica and magnesite mines in Jelšava, Lubeník and Košice. For the year 1999 the inspections in five Slovak coal mines are planned. REFERENCES [1] International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, IAEA Safety Series No. 115, Vienna, 1996 [2] 1990 Recommendations of the International Commission on Radiological Protection, ICRP Publication No. 60, Oxford, 1991 [3] Act No. 272/1994 coll., §2 [4] Protection against Radon-222 at Home and at Work, ICRP Publication No. 65, Oxford, 1993 [5] Vičanová M.; Ďurčík M.; Nikodémova D.: Sledovanie výskytu radónu v podzemných pracovných priestoroch a odhad individuálnej záťaže pracovníkov, Bezpečnosť a hygiena práce, 4,1997 21st RHD Jasná pod Chopkom SK98K0383 225 RADON EMANATION COEFFICIENTS IN SANDY SOILS K. Holý", A. Polášková0, A Baranová1*, O. Holá2), I. Sýkora1} 0 Faculty of Mathematics and Physics ofComenius University, Mlynská dolina Fl, 84215 Bratislava, Slovak Republic, 2> Faculty of Chemical Technology of Slovak Technical University, Radlinského 9, 81237 Bratislava, Slovak Republic Introduction Many authors referred that the moisture content has a significant influence on the emanation coefficient for the soils [1-3]. However, it was also found out that the effect of the water content on the radon emanation coefficient is depended on the complex of other characteristics of the soil, for example on the porosity and grain sizes [4]. Their influence on the emanation coefficient can be described only qualitatively. Therefore, in practice, the best way to obtain the information about the emanation coefficient is its measurement. In this contribution we report the results of the study of an influence of the water content on the emanation coefficient for two sandy soil samples. These samples were chosen because of the long-term continual monitoring of the Rn concentration just in such types of soils and this radon concentration showed the significant variations during a year [5]. These variations are chiefly given in connection with the soil moisture. Therefore; the determination of the dependence of the emanation coefficient of radon on the water content can help to evaluate the influence of the soil moisture variations on the variations of radon concentrations in the soil air. Materials and Methods The samples for the study of an influence of the soil moisture on the emanation coefficient were collected 2 meters from the place where the radon concentration in the soil air was continually monitored. The first sample - sandy clay - was taken from the depth of 0.1 m and the second one - sandy loam - from the depth of 1.2 m under the surface. In the first step the roughest fraction of the samples were removed. Then the samples were dried at the temperature of 105°C during several hours and subsequently their grain size analysis was done [6], The characteristics of the samples used in the experiment are listed in the Table 1. Table 1. Grain content 226Ra (%dry weight) Sample (Bq-kg1) <0.06 0.06 - 0.2 0.2 - 0.6 0.6 - 2.0 >2 (mm) (mm) (mm) (mm) (mm) Sandy loam 33.0 + 0.4 46 4 16 27 •7 Sandy clay 46.3 ±0.6 50 27 11 9 3 226 21st RHD Jasná pod Chopkom For the determination of the activity concentrations of the 226Ra the dried samples were inserted into the plastic cans of Martinelli type with the volume of 460 ml and hermetically closed. After the achievement of the radioactive equilibrium between 226Ra and 222Rn the activity concentration of 226Ra in the samples was measured by a HPGe detector placed in the low background shield [7]. The y-transitions of 351.9 keV (214Pb) and 609.3 keV (214Bi) were used for the activity determination. For the determination of the radon exhalation rates at different water contents the amount of 0.5 kg of the dried samples was steamed by the distilled water. The moisture content of the samples was determined as the difference between the weight and dry weight of the sample, divided by the dry weight. The 22Rn exhalation rates were measured by the closed chamber method. For these aims the soil sample with the volume about 0.47 / was placed into the emanation chamber of 10 /. After the radon accumulation in the tightly sealed chamber (15 - 20 hours) all radon was absorbed in the column filled with the activated charcoal and then it was transported into the Lucas cell for the 222Rn activity concentration measurement. In detail, the method is described in a previous paper [8]. Because the thickness of the soil samples was small in comparison to the diffusion length so the radon emanation coefficient could be calculated according to the relation: K=- where E is the radon exhalation rate from the sample (Bq.s'1), X is the decay constant of 222Rn (s'1), ÁRa is the 226Ra activity concentration of the sample (Bq.kg'1) and M is the dry mass of the sample (kg). Results and discussion In Fig. 1 there are shown the 222Rn emanation coefficients as a function of the moisture content. We can see that the courses are similar for both types of samples. The emanation coefficient increases with an increase of the soil moisture and after the achievement of the certain value it remains constant. It is between 5 % and 20 % of the weight moisture content for the sandy clay sample and between 8 % and 30 % of the weight moisture content for the sandy loam sample. The next increase of the moisture causes the sharp decrease of the emanation coefficient until it obtains the lower values than the measured values for the dry soil. Such courses used to be explained by the following effects in soils [9]. In the dry soil the considerable part of the radon atoms is adsorbed under the influence of the Van der Waal's forces on the internal surfaces of the soO and the emanation is relatively low. By the increase of the moisture the water hinders from the adsorption of the radon atoms on the internal surfaces of the material. Simultaneously the water presence in the internal pores of the material causes the increase of the direct recoil fraction of the emanation. On the other hand, the water presence in the internal pores reduces the diffusion of the radon atoms out of the material. However, the first two processes can be dominant up to a certain level of the moisture in the material and the increase of the emanation is occurred. After an optimum moisture content the radon diffusion is significantly reduced and the emanation is constant or it is slowly decreased. When the pores are completely filled with the water the reduced diffusion reduces the emanation dramatically. In the Fig. 1 there can be also seen that the values of the emanation coefficient are lower for the sample of the sandy loam than for the sample of the sandy clay. It is 2 ľ RHD Jasná pod Chopkom 227 obviously connected with the fact that the 38 % of weight of the sandy loam is created by the grains larger than 0.5 mm and in the sandy clay it is only 13 %. Just these grains could be responsible for the lower emanation coefficient [4]. On the contrary, almost 80 % of the weight of the sandy clay is created by the grains less than 0.25 mm for which the high values of emanation coefficient are characteristic. The ratio of the maximal emanation coefficient to that of a dry sample also gives an information about the sample composition. In case of the increase of grain size this ratio should decreases [4]. However, for both measured samples this ratio was close upon 1.3. This confirms the insignificant influence of the middle- and rough- grain's sandy fraction in sandy loam sample on this ratio. O - sandy clay • - sandy loam 0 10 20 30 40 50 Water content [% dry weight] Fig. 1. The influence of the water content on the radon emanation coefficient. The presented results show that the emanation coefficient reaches the constant value in the wide interval of the water content for both sandy soil samples. Therefore, in the common range of the soil moisture (5 % - 20 %) it is impossible to expect the variations of the radon concentration in the soil air due to the change of the emanation coefficient. The expressive changes of the radon concentration in the soil air can be observed in case of the significant decrease of the emanation coefficient during the soil drying when the water content decreases under 5 % or during the complete filling of the soil pores by the water. Acknowledgements. This study was funded by Scientific Grant Agency of Ministry of Education of Slovak Republic (VEGA project No. 1/4194/97) and International Atomic Energy Agency, Vienna, Austria (Res. Contract No. 9093/RO). 228 21" RHD Jasná pod Chopkom References [1] Tanner, A., B.: Radon Migration in the Ground: A Supplementary Review, in Natural Radiation Environment III., Symp. Proceedings, Houston, Tex. Apr. 23- 28,1978 [2] Lindmark, A., Rosen, B.: The Science of the Total Environment, 45, 397-404 (1985) [3] Goh, T., B., Oscarson, D., W., Cheslock, M., Shaykewich, C: Health Physics, Vol.61, No.3,359-365 (1991) [4] Markkanen, M., Arvela, H.: Rad. Protection Dosimetry, Vol.45, No. 1/4, 269- 272 (1992) [5] Holý, K., Bôhm, R., Polášková, A., Štelina, J., Holá, O., Sýkora, L: J. Radioanal. Nucl. Chem., Articles, Vol.209, No.2, 315-323 (1996) [6] Hyánková, A., Modlidba, L, Letko, V.: Laboratórny výskum vlastností hornín, PFUK, Bratislava, 1985 [7] Sýkora, L, Povinec, P.: Acta Phys. Univ. Comen., 31,(1990) 83 [8] Holý, K., Sýkora, L, Chudý, M., Polášková, A., Fejda, J., Holá, O.: J. Radioanal. Nucl. Chem., Letters, 199 (4), 251-263 (1995) [9] Stranden, E., Kolstad, A., K., Lind, B.: Health Physics, Vol.47, No.3, 480-484 (1984) 21" RHD Jasná pod Chopkom SK98K0384 229 RESULTS OF COMPARISON OF TWO RADON MONITORS K. Holý '\ I. Bosá1}, T. Stanys 1\ O. Holá 2), A. Polášková n, n Faculty of Mathematics and Physics ofComenius University, Mlynská dolina Fl, 84215 Bratislava, Slovak Republic, 2) Faculty of Chemical Technology of Slovak Technical University, Radlinského 9, 8123 7 Bratislava, Slovak Republic Introduction As a rule in the indoor and outdoor air the radon (222Rn) concentration is changed in a relatively short time. The measurement and study of these changes is possible by means of the continual radon monitors. But a correct calibration and verification of their operation in the environment with changing radon concentration is a precondition of their successful utilisation. Our radon monitor [1] was build at the Department of Nuclear Physics of Comenius University in the year 1990 and since the year 1991 till now it has been used for a continual monitoring of radon concentration in the outdoor atmosphere [2]. The correctness of monitor measurements has been verified by series of radon pulses of different amplitude while the radon concentration in pulses has been determined by use of the scintillation cell of Lucas type [3], The obtainment of radon monitor Alpha Guard allows for a comparison of the operation of both monitors in real conditions. Radon monitors Our radon monitor (LSCH) is build on the basis of the large scintillation chamber. The chamber has a cylindrical shape. The inner diameter is 120 mm and its length is 400 mm. The volume of the chamber is 4.5 liters. The inner space of the chamber is divided into 9 sectors to achieve good geometrical conditions for detection of radon. The walls of the sectors are coated by ZnS (Ag). Scintillations are collected through two glass windows placed at the ends of the chamber. At the flow rate of the air through the chamber of 0.5 /. min'1 and with an aerosol filter closely outside the chamber inlet the sensitivity of the scintillation chamber is 0.3 cpm at 1 Bq.m'3 of 222Rn concentration. The chamber background measured at the stable flow of the inactive air through the chamber is about 2.4 cpm which corresponds to the radon concentration of (8.3 ±1.5) Bq.m'3. The monitor allows to obtain almost 80 % of data of radon concentration in the outdoor atmosphere with an error less than 30 %. It is an adequate precision for example for the study of an influence of the changes of turbulent vertical mixing on the radon concentration in the outdoor air. A cylindric ionization chamber in a combination with DSP- technologies is used in Alpha Guard monitor (AG) for radon concentration measurement [4]. The active volume of the ionization chamber is equal to 0.56 liters. In the diffusion mode of the operation the measured gas gets via a large surface glass fiber filter into the ionization chamber. The sensitivity of the detector is 0.05 cpm at 1 Bq.m'3 of 222Rn concentration. The background signal due to an internal detector contamination is less than 1 Bq.m'3. Results and discussion For the comparison of an operation of both radon monitors the radon concentration was measured in the same unventilated room. During the measurement 230 2 ľ' RHD Jasná pod Chopkom the air was sucked with the flow rate 0.5 /.min'1 through the scintillation chamber. The radon concentrations corresponding to the time interval of 2 hours were calculated on the basis of every 30 minutes record detector's data. AG monitor operated in diffusion mode with 60 minutes lasting measuring cycle. On the basis of the recorded data the average radon concentrations corresponding to the time interval of 2 hours were also calculated similarly as for the first detector. 60 ! : : : : : 50 40 K 30 a \ v\( V{/ \ O" 20 ft TI 10 X LSC mito 0 js 50 u 40 • iD Í 1* k T. J 1 30 £ \ 'VVv vTJ \ Nl 20 / AG- man tor 1 10 0 25. 26. 27. 28. 29. 30. 31. 1. 2. 3. 4 October - November, 1997 Fig. 1. Rn activity concentration in the room measured by two radon monitors. In the Fig.l there are shown courses of the radon concentration measured by both monitors. For an illustration there was chosen such case of measurement where the course of the radon concentration was not regular during the whole measured interval. The radon concentration shows noticeable changes. The measured radon concentrations were in an interval from 11 Bq.m'3 to 53 Bq.m'3. The regular variations with the low amplitude measured in days 27. 10. - 30. 10. 1997 correspond to the high atmospheric pressure in the outdoor atmosphere (101.5 kPa) and to the regular daily variations of the temperature of the outdoor atmosphere in the range of (- 5°C, +5°C). In the period from the 31-st of October till the midday hours of the 2-nd of November the atmospheric pressure decreased and then it was at the level of 99.5 kPa till the midday of the 3-rd of November. Simultaneously the daily temperatures rose above zero and oscillated in the interval of (0°C, +12°C). In this period there is seen the interference of the regular daily variations and an increase of radon concentration in the indoor air. From the midday hours of the 3-rd of November the atmospheric pressure slowly increased and the outdoor temperature began to show the regular daily fluctuations with values less than 0°C during the night hours. It was accompanied by a decrease of the radon concentration and by an appearance of the daily variations. Further, as it can be seen in Fig.l, the courses of the radon concentration measured by means of both monitors are almost identical. Even in some details there is 2 ľ' RHD Jasná pod Chopkom 231 possible to find the same response of both monitors. For example by both devices there was measured not very expressive maximum of the radon concentration during the midday hours of the 26-th of October and also the maximum in early morning hours of 4-th of November. The average value calculated from the radon concentrations measured during the whole period is: (33.6 ± 3.9) Bq.m'3 for AG monitor and (31.1 ± 2.4) Bq.m'3 for our LSCH monitor, which is a good agreement in the scope of counting errors. 60 HI / 50 X a tio 40 30 y=A+B*x A=0,9+-l,60 20 B=l ,03-1-0,05 R=0,90 10 10 20 30 40 50 60 LSCH - 2M Rn concentration [ Bq / m 3 ) Fig. 2. Correlation plot of the 222Rn activity concentrations measured by LSCH and AG monitors. The figure 2 shows the correlation plot of radon concentrations measured by both monitors. The relation between both signals is linear and the correlation coefficient R is equal to 0.9 for the regression line. Further the constant B in regression line is close upon 1 and the constant A is close upon 0 in the scope of errors of their determination. This fact confirms again the good agreement between both monitors. Conclusion Alpha Guard monitor enabled us to test our LSCH monitor working in the environment with the changing radon concentration. The results of tests show that the records of radon concentrations of both detectors are very similar and average values of radon concentrations calculated on the basis of measured data are identical in the scope of counting errors. The values of radon concentrations given by LSCH monitor are rather more accurate as this was expected because of the larger volume of the scintillation chamber in comparison with the ionization chamber used in AG monitor. But the advantages of AG monitor are : - at the low detection limit the monitor is portable, it has the completely automated radon data collection and has also the simultaneous collection of data of the atmospheric air pressure, air temperature and air 232 2ľ'RHDJasnápodChopkom humidity. For these reasons the AG monitor is an effective instrument for the study of radon variations in the indoor air and in the case of an occurence of the increased radon concentrations also in the outdoor atmosphere. Acknowledgements. This study was funded by Scientific Grant Agency of Ministry of Education of Slovak Republic and Swiss National Science Foundation. References [1] Beláň, T., Chudý, M., Ďurana, L., GrguFa, M., Holý, K., Levaiová, D., Povinec, P., Richtáriková, M., Sivo, A. : Rare Nuclear Processes, Proc. of 14th Europhysics Conf. on Nucl. Phys., World Scientific Publishing, Singapore, 345-366 (1992) [2] Holý, K., Bôhm, R., Polášková, A., Holá, O., Sýkora, L: Results of Long-term Measurement of22 Rrt Concentration in Outdoor Atmosphere, The IRPA Reg. Symp., Prague, Czech Rep., Sept. 8-12,1997 [3] Holý, K., Sedlák, P., Beláň, T., Bôhm, R.,: The continual monitoring of the 222Rn concentrations in the outdoor atmosphere, XVI. Radiation Hygiene Days, Štrbské Pleso, Czechoslovakia, 30.11.-3.12., 1992 [4] Genrich, V.: Alpha Guard PQ 2000/ MC 50, Multiparameter Radon Monitor, Genitron Instruments GmbH, Frankfurt, Germany 21" RHD Jasná pod Chopkom SK98K0385 233 ANALYSIS OF THE AVERAGE DAILY RADON VARIATIONS IN THE SOIL AIR K. Holýl), M. Matošl), R. Bohmi), T. Stanysl), O. Holá2), A. Poláškovál) 1) Faculty of Mathematics and Physics, Comenius University, Mlynská dolina F-l 842 15 Bratislava, Slovak Republic 2)Faculty of Chemical Technology, Slovak Technical University, Radlinského 9 812 37 Bratislava, Slovak Republic Introduction In the soil air the 222Rn (radon) concentration reaches the value of several kBq.m'3 and it is not stable. The radon concentration depends mainly on the soil type, on its Ra concentration but also on the atmospheric pressure, humidity of the soil, rain precipitation and temperature [1]. Naturally, the radon concentration in the soil air presents then the result of simultaneous influence of several factors. Therefore it is difficult to search the dependence of the radon concentration in the soil on the individual factors separately in real natural conditions. The study of the short-term changes of the radon concentrations by which one of the factor is changed and the others are relatively constant is one of the possibilities to discover these influences. The example of such approach can be the search of the relation between the daily variations of the radon concentration and the regular daily oscillations of the atmospheric pressure that is also presented in this contribution. Methods The 222Rn concentration in the soil air has been continuously measured. A scintillation cell of Lucas type which volume is 125 ml has been used for counting of 222Rn alpha decays. The air was sucked from the depth of 0.8 m and dried in the refrigerator (-30° C) before entering the Lucas cell. The flow rate of the monitored air was 0.15 /.min'1. This rate was selected so that 220Rn decays before getting into the detector. The sensitivity of the radon monitor is 0.014 cpm at 1 Bq.m'3 of 222Rn concentration and its background is about 0.4 cpm. The monitor allows to measure the radon concentration in the soil air on the level about 10 kBq.m'3 with a relative error 1.5 % at one hour counting time. This precision is sufficient enough also for a study of the daily radon variations in the soil air. Results The 222Rn concentration in the soil air has been monitored continuously since 1994. The sampling place is situated in the area of Faculty of Mathematics and Physics in Bratislava only 5 m away from the sampling place for the atmospheric 222Rn measurement [2]. The soil of this place is middle permeable [3]. The mean activity concentration of 226Ra in the depth of 1.5 m is equal to 37.5 Bq.kg'1 in this soil. The emanation coefficient of 222Rn in the surface soil is approximately 14.5 % at the weight content of the soil moisture in the range of 5% - 20% [4]. The radon diffusion length is about 0.45 m in this soil and the saturated radon concentration is reached in the depth of about 1.8 m. 234 21" RHD Jasná pod Chopkom The average value of 222Rn concentration at the depth of 0.8 m determined from the three years lasted measurements is equal to 10.6 kBq.m'3. For years 1994 -1997, the average monthly values of the radon concentrations vary from 9 kBq.m"3 to 15 kBq.m'3. The measurements of the radon concentration did not show very expressive changes during a day. The amplitude of the average daily courses varies for the individual months only in an interval of (1 - 5) %. The diurnal courses of the radon concentration have not the same shapes for all months of year. However, the similar diurnal courses have been observed from May to October for all investigated years. For these months a daily minima were observed in the morning (5-11 a.m.) and the maxima in the afternoon till the evening (3-8 p.m.). Analysis Our results [5] and also the results of the other authors [6,7] showed that the regular short-term changes of the 222Rn concentration in the soil air would be connected with changes of the atmospheric pressure. According to V.I. Baranov [8] the relative change of the atmospheric pressure Apr causes the vertical shift Ah of the level with the same radon concentration AR^K) occurred in the depth h according to the relation: Ah=Apr.(H-h) (1) where His the depth of the gas permeable soil layer. Subsequently it leads to the change of the radon concentration AA&, in the depth of h because of the radon concentration increases with the depth in the under-surface layer. Provided that the radon is transported in the soil only by the diffusion, that the production rate of the radon is homogeneous and the diffusion coefficient is independent on the depth, then the depth distribution of the radon concentration for the steady state conditions is described by the relation [9]: where: AR„ (h) is the activity concentration of the 222Rn in the soil air at the depth of Ä in 3 226 3 (Bq.m" ), ARC is the activity concentration of Ra in the soil in (Bq.m' ), Ke is the 222 emanation coefficient of Rn, Fp is the total porosity of the soil, w is the volumetric soil moisture content, h is the depth beneath the soil surface in (m), L = J—— is the 222 diffusion length in (m), Def is the effective diffusion coefficient of Rn in the soil in (m2.s'') and X is the decay constant of 222Rn in (s'1). After the substitution of the equation (1) for Ah to the equation (2) we can derive the similar equation as V. P. Rudakov [10] for the change of the radon activity concentration in the depth h : . . K..A . B 1-e (3) Fp-w Because the regular daily courses of the atmospheric pressure are characterised by the low average daily amplitude below 0.1 kPa, we can modify the equation (3) to the linear form: K.Apr (4) 2 ľ RHD Jasná pod Chopkom 235 where K is the characteristic soil constant, Apr = — is the relative deviation of the Pa pressure and Ap is the deviation of the atmospheric pressure during a day from the average daily value pa. We tried to verify the relation (4) by use of the measured data. The more significant linear correlation between Mb, and Apr was obtained predominantly for the summer months from May to August when the soil surfece was dry (the correlation coefficient R e < - 0.7, -0.9 >. The mean value of the soil constant K = (- 450 ± 80 ) kBq.m-3 was determined from the correlation plot of Mn„ and Ap, for data from July 1995, August 1996 and August 1997. For our parameters of the soil (Ke, L, AR, , w, Fp) this value of K corresponds to the depth of the gas permeable soil layer of H= (44 ± 9) m, which can answer to the reality. In regard to these results we can modify the equation (4) by such a way in the following step that we substitute instead of Ap the well known relation describing the changes of the -99.45 11.2- -99.25 O 2 4 6 8 10 12 14 16 18 20 22 24 Time [hour] The average daily course of the radon activity concentration in the soil ( —•— measured radon activity concentration, — radon activity calculated according to the equation (5), -o— measured atmospheric pressure). atmospheric pressure during a day [11]. Finally, we obtain the following relation for the daily course of the radon activity concentration of 222Rn in the depth h: (5) — W pa p pa 222 where A^m (h) is the average daily value of the activity concentration of Rn in the depth h, a\ , a% , fa , fa are the amplitudes and phase angles, respectively, of the first and second harmonic terms of the daily course of the atmospheric pressure, t is the time 236 2ŕRHDJasnápodChopkom from the beginning of the day in hours. The meaning of the other symbols is the same as in the previous part. We applied the relation (5) to the description of the average daily courses of the radon activity concentration for the individual months. In the Fig. 1 there is illustrated that the average daily course of the radon activity concentration in the soil air can be very well described by the equation (5) for the summer months. The agreement between the calculated and measured average daily courses of the radon activity concentration is very good. The harmonic analysis of the average daily courses of the radon activity concentration for July 1997 also showed that the amplitude of the first harmonic term is three times higher than the amplitude of the second harmonic term. Therefore, in this period the thermal action on the daily course of the 222Rn concentration is dominant. Conclusion The deviation of the radon activity concentration in the soil air from the average daily value reaches only a few percent. For the dry summer months the average daily course of the radon activity concentration can be described by the equation (5). As it can be seen in the equation (5) the analysis of the average daily courses could give the information concerning the depth of the gas permeable soil layer. This soil parameter is determined by others methods with difficulty. Acknowledgements. This study was funded by Scientific Grant Agency of Ministry of Education of Slovak Republic and International Atomic Energy Agency, Vienna, Austria (Res. Contract No: 9093/RO). References [I] Taipale, T., T., Winquist, K.: The Science of the Total Environment, 45, 121- 1269 (1985) [2] Holý, K., Bôhm, R., Polášková, A., Holá, O.: Results of Long-term Measurement of 222Rn concentration in Outdoor Atmosphere, The IRPA Reg. Symp., Prague, Czech Rep., Sept. 8-12,1997 [3] Klasifikace zemin pro zakládání staveb. ČSN 73 1001 [4] Holý, K., Polášková, A., Baranová, A., Holá, O., Sýkora, I.: Radon emanation coefficients in sandy soils. XXI. Radiation Hygiene Days, Jasná pod Chopkom, 23.-27.11. 1998, Slovakia [5] Holý, K., Bôhm, R., Matoš, M., Polášková, A., Holá, O.: Study of 222Rn variations in the soil air (poster). The IRPA Reg. Symp., Prague, Czech Republic, September 8-12,1997 [6] Tretjakova, S., O., Džolos, L., V., Neresov, L, L., Vojtov, G., L, Pavlov, V., D.: Report JINR, Dubna, 18-83-445,1983, p.l [7] Harley, N., H., Chittaporn, P.: 40 th Annual Meeting of the Health Physics Society, 23- 27 July 1995, Boston, Massachusetts, Abstract THAM-A4 [8] Baranov, V., L: Radiometrija, Moskva, Izd. AN SSSR, 1995 [9] Dôrr, H., Munnich, K., 0.: Tellus, 42 B, 20-28 (1990) [10] Rudakov, V., P.: Fizika Zemli, No.7,124-127 (1985) II1] Tomlain, J. in: Klíma a bioMíma Bratislavy, Veda, Bratislava, 1979, p.71 2ŕ RHD Jasná pod Chopkom SK98K0386 237 THE SURVEY OF DWELLINGS WITH INCREASED RADON LEVELS IN SLOVAKIA Magdaléna Vičanová Institute of Preventive and Clinical Medicine Limbová 14, 833 01 Bratislava Introduction The Slovak national survey of indoor radon exposure started in the early of 1990s, when legislative provisions were prepared for protection of the population from radon and its daughter products. In 1992 the Slovak Ministry of Health published an ordinance about the protection of population to radon and other natural radioactive sources. The action level (A.L.) of annual average equilibrium equivalent concentration (EEC) of radon was established as 200 Bq.m'3 for existing dwellings and 100 Bq.m"3 for built in the future. This national survey of indoor radon measurements in a sample of dwellings in Slovakia was organised by the Institute of Preventive and Clinical Medicine (IPCM) in Bratislava. The aim was to find districts and type of dwellings with the highest indoor radon concentrations and to estimate the radiation load of the Slovak population owing to indoor radon exposure. Materials and Methods Passive solid state nuclear track detectors were used to measure indoor radon concentrations. The detectors were polyallyldiglicolcarbonate CR-39, which were placed in about 6,000 selected houses (minimum two detectors for every residence) and distributed by professional workers from regional Specialised State Institutes of Public Health. Questionnaires for obtaining the detailed information required for interpretation of results were distributed to each house. After six months exposed detectors and questionnaires were returned to IPCM for analysis. Electrochemical etching combined with a chemical pre-etching process was used for evaluating detectors. Quality control calibration of detectors was carried out in the reference radon and radon daughter measuring chamber, at the State Metrological Centre of the Institute of Preventive and Clinical Medicine. Results of indoor radon measurements with questionnaire data are stored in a special database program. The software provides the possibility comparison of the measured radon concentrations with reference to building type and other parameters. Experimental Results Our present results are from 3,657 residences (0,2% of total dwellings in Slovakia) . It was found that the arithmetic mean (AM) of EEC was 86±119 Bq.m "3, the geometric mean (GM) was about 41±2.22 Bq.m "3 and 11% of dwellings (N=409) have a greater EEC of radon than the action level. 238 21"RHD Jasná pod Chopkom The arithmetic and geometrie mean of EEC varied between type of building and districts of Slovakia. The sample of family houses (N=2,363) has AM 125±135 Bq.m"3, GM 73±1.8Bq.m and the sample of multifamily houses (N=l,294) has AM 22±24Bq.m"3, GM 15±1.46 Bq.m"3. We calculated the population-weighted AM of EEC for every district by different type of house, and then estimated this value for the whole of Slovakia obtaining a figure of 48 Bq.m'3. Because 407 family dwellings and only 2 multifamily dwellings have a greater EEC of radon than the action level we will target to only family dwellings. Their distribution according to year of building is in the table 1. Tab.l Distribution of family dwellings according to year of building. Year Number No of dwellings No of dwellings of build of dwellings with EEC > A.L. with EEC > A.L. > 1979 532 34 6.4 1960 -1979 886 142 16.0 1940 -1959 352 95 27.0 1920 -1939 179 55 30.7 1900-1919 68 29 42.6 < 1899 63 23 36.5 not know 283 29 10.2 Our results from family dwellings show, that those with cellars (when cellars are below whole building) have lower radon levels (N=827, AM 77 ±109 Bq.m'3, GM 46±1.72 Bq.m"3) than those without cellars (N=601, AM 146±138 Bq.m'3, GM 97±1.61 Bq.m'3) or which have cellars only below part of building (N=737, AM 148±148 Bq.m , GM 93±1.77 Bq.m"3). The same is true when we look at the results for ground-floor rooms. Rooms above cellars have lower radon levels (N=l,206, AM 74±108 Bq.m'3, GM 40±1.91 Bq.m'3) than rooms directly above ground (N=1.341, AM 134±137 Bq.m"3, GM 82±1.78 Bq.m'3). Table 2 shows distribution of radon levels in the rooms according to the floor. The highest radon levels are in the basement or ground- floor rooms. Tab.2 Distribution of EEC of radon Place Number AM GM of rooms [Bq.in'3] [Bq.ni"3] Basement 133 106±138 61±1.82 Ground-floor 3,337 107±127 61±1.93 First floor 1,118 67±73 43±1.65 Conclusions The national survey results suggest that Slovakia may be among the countries with high radon risk in Central Europe. The population-weighted arithmetic mean is 48Bq.m"3, the maximum value found was 1500 Bq.rn'3 and the average annual effective dose from indoor radon exposure is 2.1 mSv. The districts with the highest indoor radon 21" RHD Jasná pod Chopkom 239 concentrations correlate with known presence of uranium in the soil, therefore the soil is probably the main source of radon in Slovak dwellings. Our survey of dwellings with increased radon levels supported this conclusion, because the highest radon levels were found in older family houses without cellars. Acknowledgements We are grateful to all professional workers from regional Specialized State Institutes of Public Health for their help at distribution radon detectors, we thank Mrs.Oľga Mlynárova and Mrs.Larisa Kulková for their excellent technical help. References 1. Vičanová M., Ďurčík M., Nikodémova D.:"Indoor Radon Exposure of Slovak Population." European Conference Protection against Radon at Home and at Work, Book of Proceedings, Prague, June 1997 2. Vičanová M..Nikodémova D., Ďurčík M.,Havlík F.: Indoor Radon Concentrations in "Hot Spots" locations in Slovakia. Proceedings of Healthy Buildings '95, Miláno 10- 14 September 1995, Vol.2, po.715-720 3. F.Havlík, M.Ďurčík and D.Nikodemová, State Metrological Centre of Slovakia for Radon Quantities, Bezpečnost jadrovej energie, 1(39),1994, No.4. 4. UNSCEAR-Report 1993, Sources and Effects of Ionizing Radiation, United Nations, New York 1993. 5. Inter. Basic Safety Standards for Protection against Ionising Radiation, Saf. Ser. 115, IAEA Vienna, 1996 240 SK98K0387 21"RHD Jasná podChopkom VARIATIONS OF RADON VOLUME ACTIVITIES IN SOIL AND INDOOR AIR AND THEIR CORRELATION Andrej Mojzeš Department of Applied and Environmental Geophysics, Faculty of Natural Sciences, Comenius University, Mlynská dolina - G, 842 15 Bratislava, Slovakia Introduction If we neglected the contribution to the 222^ volume activity (VAR) in indoor air originating from building materials, which is really small for commonly used building materials in Slovakia [7], the main cause of indoor radon presence would be its release from underlying rocks of house. The exhalation of radon from rock environment to atmosphere is a complex process and this complexity is multiplied by an existence of building as an object on the boundary of these diametrically different spheres. Several groups of parameters determine this process : a) parameters of geological basement : mineralogical composition and content of parent radioelements (uranium, radium, thorium) in rock and its soil profile, structure and texture of rock and soil, porosity and gas- and water permeability of soil in the case of quasi-homogeneous rock environment; parameters of faulted zone - tectonics (width, length, slope, intensity of depth' communication) in the case of disjunctively disturbed rock environment which actually appears much more frequently as it is commonly noticeable and acceptable, b) parameters of atmospheric environment : temperature, pressure, humidity, velocity and direction of wind, precipitation, inversions, etc. whose changes could have their reflection in pore space of rocks and soils to different depth and could serve as one of moving mechanisms [5] of radon exhalation (temperature, pressure, humidity gradient, etc.). For that reason it is essential to know the nature and the reasons of time- dependent variations of radon volume activities for different rock formations, c) parameters of specific building (depth of foundations, presence of cellar, technique of insulation from basement soil, height of house, chimney, method of ventilation, age, ...) which through its presence considerably affects "standard" interaction between geological bedrock and atmosphere [4], d) parameters of interaction litosphere - atmosphere (ionosphere). The exchange of huge amounts of matter in molecular and ionic form between these spheres of Earth mainly in areas of faulted zones in the Earth crust which could be identified by radon as relatively easy measurable element [1,2]. Characteristic of measurements Some manual measurements of volume activity of 222^ jn ^j] gfc ^d in indoor air of building together with parallel measurements of some meteorological parameters (temperature, humidity and pressure) of both atmospheric and indoor air were carried out. The measurements were performed in the building of Faculty and in its subsoil which consists of slope loams at the base of SW slopes of the granitic Malé Karpaty Mts. in the area of confluence of the Vidrica Creek with an arm of the Donau River. The monitoring measurements lasted for more than one and a half year, from January 1997 to August 1998, with the frequency of approximately once a week in each object. 21st RHD Jasná pod Chopkom 241 The soil air was taken from a permanently set up and sealed pipe from the depth of 0.8 m which was placed approximately 10 m from the building at the open air. The work [3] which presents some results of soil and atmospheric radon monitoring measurements at the point at a distance of approximately 100 m, gives these features of subsoil: the soil is fine-grained, the content of fine-grained particles with diameters d < 60-10"6 m is about 68 %, the mass activity of 226Ra = 37.5 Bq/kg, the total porosity = 0.55, the dry bulk density = 1090 kg/m3 , the coefficient of emanation = 14.5 % and the volumetric soil moisture content = 19 %. The building of Faculty is a typical reinforced-concrete building with large windows' surface without air conditioning but with very small sealing-off and therefore with poor insulation from outdoor environment. The measurements in the building were carried out in one room in the basement and in one room on the 3-rd floor for the purpose to verify an influence of distance from radon source. No one of rooms was specially sealed for this purpose. We only fulfilled closing of windows and turning the heating off in the basement room but there was a more free regime in the room on the 3- rd floor, the room was occasionally aired in the summer and the heating worked in the winter. All measurements of 222RJJ volume activities were performed with a portable fully automatic scintillation detector based on exchangeable Lucas cells (LUK 3R). Since the instrument is primarily designed for soil 222Rn assessment our indoor radon volume activities are not reliable in their absolute values but we managed to record their relative changes and this could be sufficient for our purposes. There were also performed the parallel measurements of some meteorological parameters (temperature, humidity and pressure) of air in each object. Results of measurements The results of measurements are presented in Fig. 1. The graphs show the running averages of the volume activities of both soil 222jjn [kBq/m3] (513 observations) and indoor 222Rn [Bq/m3] (281 + 252 observations in both rooms) together with their mean values (average) and the running averages of the observed temperatures, humidities and pressures of air during the measured period. The geological basement of building is a source of indoor radon. The volume 3 3 activities of soil 222]^ range from about 2 kBq/m to about 20 kBq/m with the average of 9.26 kBq/m3 and the standard deviation of 2.95 kBq/m3. The course of the volume activity of soil radon is relatively in good agreement with the results of mentioned work [3] mainly as concerns the maximum values measured in summer months (June - August) which are connected with increased rainfall, typical for this area in this season. The extra expressive maximum in period July - August 1997 coincides with very strong rainfall (inundations on the Morava and Váh Rivers). The second maximum usually located in winter months [3, 6] found smaller expression in our results. The dependence on distance from the source was confirmed in indoor radon measurements : the volume activities of 222RU -m T0Qm On the 3-rd floor are generally lower than those ones in basement room, i.e. closer to the source. The visual correlation between the courses of indoor radon in both rooms and the course of soil radon is quite different. This is related either to the distance from the basement or , as mentioned, to the different ventilation system of those rooms. As it can to be seen in Fig. 1 the course 242 2 ľ1 RHD Jasná pod Chopkom Fig.l Volume activities of 222Rn in soil and indoor air and some atmospheric and indoor air parameters 200-, INDOOR AIR on the 3-rd story -L-920 i í MNI I I I I ! t I I LEGEND: volume activity of radon SOIL AIR mean value of volume activity of radon at the depth of 0.8 m air tempe raturefC] ať humidity [Kr.h.l barometric pressure IhPa] 2 ľ RHD Jasná pod Chopkom 243 of volume activities of 222Rn in the room in basement falls in very good with the course of volume activities of 222Rn in the basement soil. This fact gives an evidence of an excellent and fast communication based on direct contact between the subsoil and the air interior of basement rooms. The correlation between the course of volume activities of 222Rn in the room on the 3-rd floor and the course of subsoil radon is substantially lower even if some maxima on the VAR curve in the room could have their explanation in time delay after the maxima on the curve of soil radon (e.g. the expressive maximum of indoor VAR on the 3-rd floor approximately in August 1997 could be a consequence of the maximum of soil radon in the period June - August 1997, ditto in summer 1998). On the contrary, the most expressive maximum of indoor VAR on the 3-rd floor is in the-period November 1997 - January 1998 as a result of better airtightness of room because of heating season. We tried to support the visual correlation or non-correlation between the measured courses of volume activities of 222Rn in subsoil and in study rooms with a calculation of coefficients of correlation by the method of simple linear regression. The results are presented in Tab. 1 which shows the tested variables, parameters of regression y = kx + q, coefficient of correlation r and number of analysed observations N. Mainly the coefficient of correlation between the soil VAR and the indoor VAR in the basement room is apparently inconsistent with the good visual correspondence of the courses of both quantities, because of its small value. There are also presented the correlation relationships calculated between the VARs and the measured meteorological factors in each object in Tab. 1. There are positive correlations between VARs and air temperatures for the basement soil and the basement room which are caused by an existence of the most expressive VAR maxima in summer seasons. The equal values of their coefficients of correlation could also indicate the good communication of these objects. The most expressive positive correlation between the VAR in soil and the atmospheric air humidity is connected with an accumulation of soil air below the water saturated upper layer of soil during a rainfall period which results in decreasing of gas permeability of soil and therefore also in decreasing of radon exhalation through the ground surface into the atmosphere. The higher water saturation of soil (rainfall intensity and duration) the higher VAR amplitude in soil air. The value of the coefficient of correlation between the VAR in basement room and the humidity of its air is lower because of indoor air measurement. The relationships between the indoor VAR on the 3-rd floor and both the temperature and humidity of indoor air are considerably deformed because of heating season and lower airtightness. An influence of barometric pressure on the VAR in each study object is different. Whereas its influence on the VAR in soil air at the depth of 0.8 m is generally negligible it seems, based on the coefficients of correlation for the rooms, that the increasing of barometric pressure causes the accumulation of radon in basement parts of building, on the contrary, its decreasing leads to "sucking" of radon to upper parts of building. Conclusions The results of monitoring measurements during 20 months' period point out the intensity of interaction of geological substrate with building interior through the values of the volume activity of 222Rn. Therefore a method of building foundation is one of the most important actors which determines the quantity of radon in indoor air. In the light of quality, the fluctuation of radon presence in the bottom parts of the building is 244 2 ŕ RHD Jasná pod Chopkom strongly determined by the variations of soil radon, this course is controlled by the ventilation and heating system on upper floors. Tab.l Variables x,y tested by linear regression y = kx + q, coefficient of their correlation r and number of observations N DEPENDEN INDEPENDENT k q r N T VARIABLE X VARIABLE Y VAR in VAR in soil 1.06 164.7 0.08 281 basement room VAR in room VAR in soil 2.78E-5 147.24 3.21E- 208 on the 3-rd 6 floor VAR in soil temp. of atm. air 0.08 8.93 0.22 400 SOIL VAR in soil humidity of atm. 0.07 7 0.29 390 air VAR in soil -0.01 20.03 -0.03 382 pressure of atm. air VAR temp, of indoor 1.16 155.38 0.22 238 ROOM air IN VAR 0.34 162.22 0.11 238 BASEME humidity of -. NT : VAR indoor air 1.12 0.23 222 pressure of indoor 888.53 air VAR temp, of indoor -0.44 158.63 -0.06 209 ROOM VAR air -0.09 151.3 -0.03 209 ON THE humidity of 3-rd VAR indoor air -0.29 426.71 -0.11 195 FLOOR pressure of indoor air References [1] GRUNTORÁD J.: Dynamický model přirozeného atmogeochemického pole, jeho měření a vztah k životnímu prostředí [Dynamic Model of Natural Atmogeochemical Field, Its Measurement and Relationship to Environment]. In: Zb. abstr. zo semináru "Atmogeochemické pole - nové poznatky ve vztahu k ŽP, zejména k problematice radonu", Silikátová společnost, Prague, 1997,1-2 [2] GRUNTORÁD J., MAZÁČ O.: Impact of Subtle Dynamic Geofactors on Environment. Acta Universitatis Carolinae Environmentalica, vol.8 (1995), 3-53 [3] HOLÝ K., BÔHM R., POLÁŠKOVÁ A., ŠTELINA J.: Variations of 222Rn concentration in outdoor atmosphere and in soil air. In: Conference Proceedings "19"1 Radiation Hygiene Days", Jasná pod Chopkom, 1995, 129-131 2ľ1 RHD Jasná pod Chopkom 245 [4] JIRÁNEK M., POSPÍŠIL S.: Radon a dům [Radon and House], ARCH, Prague, 1993,17-41 [5] KRISTIANSSON K., MALMQVIST L.: Evidence for nondiffusive transport of 222Rngg ju the ground and a new physical model for the transport. Geophysics, vol.47 (1982), 10,1444-1452 [6] MATOLÍN M., PROKOP P.: Variation of radon volume activity in soil air in a year climatic cycle. In: Radon investigations in Czechoslovakia in., Czech Geological Survey, Prague, 1992,1-5 [7] NIKODÉMOVA D., VIČANOVÁ M., HAVLÍK F., ĎURČÍK M.: Radón a zdravie obyvateľstva [Radon and Public Health], ÚZV, Bratislava, 1995, 13-17 246 SK98K0388 21"RHD Jasná pod Chopkom THE POSSIBILITY OF EFFECTIVE AND EQUIVALENT DOSE DETERMINATION EV THE NATIONAL PERSONNEL DOSIMETRY SERVICE J.Trousil, J.Plichta, J.Síudená, J.Štrba CSOD Praha, ČR Abstract The paper is describing the process of E and HT determination from responses of film, TL and neutron track dosimeters used in CSOD for monitoring of persons working under a radiation risk. The responses of mentioned dosimeters - optical density, integral of TL glow curve and track density are transferred to apparent doses though a routine calibration and by means of determined photon and electron energy and conversion coefficients (ICRP74) the apparent doses are transferred to Hp(10) and Hp(0.07). As for the film badge method, the angle of radiation incidence is also taken into account. Routine calibration is realised by the quantity Ka, the determination of quantities Hp(10) and Hp(0.07) requires a complex calibration on a phantom in the energy range between 10 keV - 6 MeV and an application of conversion coefficients Hp(10)/ Ka or Hp(0.07)/Ka (ICRP74). The quantity Hp(10) is used for the calculation of E by means of determined radiation energy and corresponding conversion coefficients E/ Hp(10) from ICRP74. The Hp(0.07) quantity is used directly as HT on the skin. In the case of neutron dosimetry, the track density is transferred by means of routine calibration with Cf-252 to an apparent dose and by means of determined neutron energy and conversion coefficient Hp(10)/O (ICRP74) the quantity Hp(10) is calculated. The quantity E is calculated from Hp(10) and conversion coefficient E/Hp(10) from ICRP74. The quantity is presented in the annual dose reports. 2 ľ' RHD Jasná pod Chopkom 247 AUTHOR INDEX Adámek P. 179 Jarchovský D. 33 Auxtová Ľ. 149,156, 179 Jirsa L. 82 JudasL. 113 Baranová A. 225 Jurchová B. 24 Bečková V. 139 JurzaP. 183 Beňo M. 102 Blažek T. 82 Kaizer J. 28 BôhmR. 213,233 Kárny M. 82 Borovicová F. 67 Kassai Z. 174 Bosá I. 213,229 Kavan P. 72 KlisákJ. 167 Cabáneková H. 144, 205 Klusoň J. 183 Čechák T. 87, 183 KoprdaV. 174 Češpírová I. 135 Korec J. 107 Ciubotaru A. 140 Koskelo M. 187 Constantin F. 140 KrnáčŠ. 159,187,192 Compel i 164 Kroutilíková D. 113 Kuruc J. 37 Dillinger P. 59, 61,152 Dobiš Ľ. 28 Letkovičová M. 107 Došel P. 72 LukáčP. 59,61 ĎurčíkM. 121,201,209 Ďurec F. 149,156,179 Malátová I. 56,135,139, 167 Ďurecová A. 149,156 Matoš M. 233 Mihály B. 107 Fiala E. 24 Mojzeš A. 240 Filip J. 24 Foltánová Š. 139, 167 Novotný J. 87 Fôldesová M. 59,61 Novotný J. Jr. 87 FtáčnikováS. 47,76 Nikodémova D. 52, 92, 96,102, Furiová A. 67 121,209 Futas M. 222 Nimec. J 82 Gombala E. 149,156,222 PatzeltováN. 61 Petrášová M. 52 Harangozó M. 152,174 Petrová K. 63,128 Heřmanská J. 72,82 PetrufP. 217 HoláO. 213,225, 229, Plichta J. 246 233 Polášková A. 213,225,229, Holý K. 213,217,225, 233 229,233 Popa D. 140 Horváthova M. 92,96 Prikazský V. 107 Hušak V. 128 ProuzaZ 16,63 Husár J. 67 248 21" RHD Jasná pod Chopkom Ranogajec M. 92,96 ŘehákR. 107 Rulík P. 56 ŠeligaM. 42 SpéváčekV. 87 Spurný F. 130 Stanys T. 217,229,233 Starostová V. 36 Štrba J. 246 Studená J. 246 Švec A. 125 Svitek J. 28 Sýkora I. 225 Thomas J. 197 Tintěra J. 87 Tô'lgyessy J. 152 Tomášek L. 97 Tomášek M. 56,171 Trousil J. 246 Venkatamaran R. 187 Vičanová M. 102, 201,209, 237 Vladár M. 11,92,96,102, 144,205 Vošmiková K. 82 Vymazal J. 87 Žáčkova D. 113 Zelenka Z. 20,24 Zimák J. 72,82 AUTOMATIC PERSONAL INTERACTIVE DOSECONTROL TLD AREA MONITORING ADAPTIVE DOSIMETER SOFTWARE DOSIMETRY SYSTEMS DOSIMFTERS READERS PACKAGES SYSTEMS WASTE LAUNDRY AND BODY PORTAL GAMMASCAN MEASUREMENT SMALL ITEMS CONTAMINATION CONTAMINATION VEHICLE SYSTEMS MONITOR MONITORS MONITORS MONITOR FINLAND GERMANY ITALY USA UK PADOS Technology Oy RADOS Technology GmbH RAOOS Technology S.r.l. RADOS Technology Inc. RADOS Technology Ud. P O. Box 506 P.O. Box 501245 Vlale Parioli, 47 6460 Dobbin Road Crown House, Kings Road West. Newbury FI."f.20f 01 Turku D-22712 Hamburg 00197 ROMA, Italy Columbia, MD 21045 Berkshire RS14 5BV Tol.+358-2-468 4600 Tel.+49-40-85193-O Tel.+39-6-412 2171 Tel. +1-410-740-1440 Tel.+44-1635-49429 F.K.+358-2-4684601 Fax.+49-40-85193-256 Fax.+39-6-41221707 Fax.+1-410-740-4676 Fax. +44-1635-37335 Representatives in more than 30 countries. •-.!• • i • 7 i i i 71 H.• u j , AUTOMATIC PERSONAL INTERACTIVE DOSECONTROL TLD AREA MONITORING ADAPTIVE DOSIMETER SOFTWARE DOSIMETRY SYSTEMS DOSIMETERS READERS PACKAGES SYSTEMS WASTE LAUNDRY AND BODY PORTAL GAMMASCAN MEASUREMENT SMALL ITEMS CONTAMINATION CONTAMINATION VEHICLE SYSTEMS MONITOR MONITORS MONITORS MONITOR /ustúpte pn> C eskou republiku: Metrxcz si o ^VJ89 130C3P-dha3 til iO2)-173 2161 71754E02 ET l-w (02i6731 1'25 i -nnf n<* níľnilio. TEXT CORRECTION TO: RESULTS OF RADIATION PROTECTION AT NUCLEAR POWER PLANT DUKOVANY 1988 -1997 Zdenek Zelenka ČEZ, Inc. - Nuclear Power Plant Dukovany Table 2: Distribution of Number of Workers by Annual IDE interval Annual IDE interval [mSv] Year <0.1 0.1-0.5 0.5-1 1-2 2-5 5-10 10-15 15-20 20-25 25-30 1991 959 860 182 134 117 10 0 0 0 0 1992 869 880 180 137 168 48 4 0 1 0 1993 1121 934 184 160 108 55 9 1 1 1 1994 1659 454 140 125 94 30 6 5 2 1 1995 1579 422 169 150 111 34 13 10 1 0 1996 1507 426 145 154 105 53 6 0 0 0 1997 1345 494 179 146 135 32 3 4 1 0 1.37x10* D «,«, [mGyl 40.78 5 T ABo,v[Bqsi< > 3x10" Br^/DT^fSv/Gy] 3.02 Sr^Bod, [Gy/Bq si(>) 1.63x10" Sra^T^AuBrt [Gy/Bq s]m 6.36xlO14 STe,,«<.T«t«(oii..r) [Gy/Bq s](Z) 1.69xlO13 C)MIRDOSE3.1