IAEA-TECDOC-750
Interim guidance for the safe transport of reprocessed uranium
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INTERIM GUIDANC SAFE TH E R TRANSPOREFO REPROCESSEF TO D URANIUM IAEA, VIENNA, 1994 IAEA-TECDOC-750 ISSN 1011-4289 Printe IAEe th AustriAn y i d b a June 1994 FOREWORD
Increasingly reprocessed uraniu ms beini e gfabricatio th use r fo d f nucleao n r fuel elements. This entails chemical separatio uraniue th f no m isotopes, conversio uraniuo nt m hexafluoride, re-enrichment, reconversion to uranium dioxide and fuel fabrication. Since man thesf yo e processes take plac t differenea t locations, transpor uraniuf o t m material will be necessary. Because of the slightly different isotopic composition of reprocessed uranium and the presence of traces of impurities, the radiation levels are generally higher than for unirradiated uranium. These different properties cast some doubt on the presumption that packages used for the transport of unirradiated uranium are automatically suitable for the transport of reprocessed uranium compounds.
IAEe Th A Transport Regulations themselve rathee sar r ambiguou thin so s matter, since at the time that these were drafted, transport of reprocessed uranium compounds was a rare event. The Standing Advisory Group on the Safe Transport of Radioactive Material (SAGSTRAM) recommended thaissue th treviewe e eb consultanty db thad documensan a t t developee b d that would give guidanc Regulations e usero et th f so .
This TECDOC is the result of the endeavours of the experts convened at two Consultants Services meetings t I contain. s guidanc e provisionth e currenn th o e n i s t Regulation wels sa proposals a l Revisew changer ne sfo e dth Editioo st n whose publication is planned for 1996. This document demonstrates that under the present Transport Regulations it is possible in most cases to ship reprocessed uranium compounds in the same packages as unirradiated uranium compounds. In few cases a more stringent package type is required. It is expected that this document will assist to clear out any uncertainties on the correct interpretatio e presenth f o nt Transport Regulation o ensurt d ean s international acceptance on the conditions for transport of reprocessed uranium.
Further information on the transport of reprocessed uranium may be obtained from H.A. Selling, IAEA Division of Nuclear Safety. EDITORIAL NOTE
In preparing this document for press, staff of the IAEA have made up the pages from the original manuscript(s). viewsThe expressed necessarilynot do reflect those governments ofthe ofthe nominating Member States nominating ofthe or organizations. The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions delimitationthe of or of their boundaries. The mention of names of specific companies productsor (whether indicatednot or registered)as does implyintentionnot any infringeto proprietary rights, should construednor be it an as endorsement recommendationor partthe IAEA. ofon the CONTENTS 1. INTRODUCTION ...... 7 . 2. EXPLANATORY NOTE ON REPROCESSED URANIUM TRANSPORT ..... 8 2.1. Reprocessed uranium subjec transporo t t t ...... 8 . 2.2. Characteristic reprocessef so d uranium ...... 8 . . DEFINITIO3 UNIRRADIATEF NO D URANIUM ...... 0 1 . 4. APPLICATION OF THE REGULATIONS TO THE TRANSPORT OF REPROCESSED URANIUM ...... 1 1 . Regulationse 4.1Th ...... 1 1 . 4.2. Application of the Regulations ...... 11 4.2.1. Principles of the Regulations ...... 11 4.2.2. Reprocessed uranium classifie LSA-Is da I ...... 2 1 . 4.2.3. Reprocessed uranium classified as LSA-III ...... 12 4.2.4. Consideration of the daughter nuclides of uranium ...... 14 4.2.5. Classificatio enrichef no d commercial grade uranium ...... 4 1 . 4.2.6. Observation radioactivn so e impurities ...... 4 1 . 4.2.7. Consideratio changef no dosn si e equivalent rate with time .....4 1 . 4.2.8. Handling of cylinders emptied of uranium hexafluoride ...... 15 4.2.9. Additional requirement for an IP type package containing reprocessed uranium ...... 15 4.2.10. Approval by the authorities and multilateral approval of reprocessed uranium transport ...... 15 4.3. Marking of packages and inspection before shipment ...... 15 5. SUMMARY ...... 7 1 . APPENDIX I: THE RADIOACTIVITY OF URANIUM ...... 19 APPENDIX II: NUCLIDES COMPOSING REPROCESSED URANIUM ..... 21 APPENDIX III: IAEA REGULATIONS CONCERNING REPROCESSED URANIUM ...... 23 APPENDIX IV: DEFINITIONS OF UNIRRADIATED/REPROCESSED URANIUM AND THEIR IMPLICATIONS ...... 25 APPENDIX V: REPROCESSED URANIUM CLASSIFIED AS LSA-II ...... 33 APPENDI : XMEASUREMENVI LEACHINF TO G RATI URANIUF OO M OXIDES ...... 7 3 . APPENDIX VII: TREATMEN URANIUF TO M POWDERE TH N SI REGULATIONS ...... 43 APPENDI XMIXTURF VIIIO 2 A : E ...... 5 4 . APPENDI VALU2 A URANIU: F EXO IX M WHIC SUSPECTEHS I HAVO DT E BEEN CONTAMINATED BY REPROCESSED URANIUM ... 47 APPENDI : XX CONTRIBUTIO RADIOACTIVF NO E IMPURITIEN SI REPROCESSED URANIUM TO THE VALUE OF A2 ...... 51 APPENDI : XDAUGHTEXI R NUCLIDE THEI 228D F STO hAN R INGROWTH ...... 55 APPENDIX XII: CLASSIFICATION OF AN EMPTY CYLINDER CONTAINING UF6 HEEL ...... 57
ANNEX: INVESTIGATIONY TRANSPORE 48 N TH I 6 N SO UF F TO CONTAINERS WITH URANIUM FROM REPROCESSED PWR FUEL ELEMENTS ...... 59
REFERENCES ...... 3 7 .
CONTRIBUTOR DRAFTINO ST REVIED GAN W ...... 5 7 . 1. INTRODUCTION
Applicatio Regulationse th f no (1985 Edition transpore th o )t reprocessef o t d uranium has been under examination followin Meetinh e 7t Standine decisio th e th g f th go y gnb Advisory GrouSafee th n pTransporo Radioactivf o t e Material (SAGSTRAM IAEAe th f )o .
decides Technicae wa th t I t da l Committee Meetin Continuoue th r gfo s Reviee th f wo IAEA Regulations for the Safe Transport of Radioactive Materials and the Supporting Documents (TCM-405.3, 12-16 July 1989) to enlist consultants services for examination of followine th g items concerning transpor reprocessef o t d uranium:
(1) To develop proposals for provisions suitable for the incorporation of reprocessed uranium in the next edition of Safety Series No. 6 [1], especially in the definition of uraniu 150d an paramn i .9 s14
(2) To find out the radiological effect caused by the real composition of reprocessed uranium (e.g. other uranium isotopes, transuranic elements, fission products, daughter product othed san r impurities) taking into account practical problems likbuildue eth p of daughter product 'heels' in UF6 cylinders recycled many times.
Consultant services meetings (CSMs) were hel Novemben di r 198 Marcd 9an h 199o 2t examine the above issues. This "Interim Guidance for the Safe Transport of Reprocessed Uranium" summarizes the results of these two meetings.
In this document, names of certain IAEA publications are abbreviated as follows:
(1) Safety Series No. 6 [1]; The Regulations.
(2) Safety Series No. 7 [2]; 557.
(3) Safety [3]7 Serie3 . ; SS37.sNo
Conformit transpore th f yo reprocessef to d uranium wit Regulationse hth examines wa d to determine whether packages used for the transport of unirradiated (enriched) uranium may als usee or b reprocesse dfo d uranium transport resula s f thiA o .t s scrutiny s beeha nt i , concluded that the provisions in the current Regulations that govern transport of natural uranium can generally be applied to reprocessed uranium compounds and that reprocessed uraniu consequentln mca transportee yb specifiw lo a cs d a activit y (LSA) material.
However e revisio e Regulationsth , th f no themselve s unde i swity wa ha vier o wt completio 1996n ni . This document therefor interin a s ei m document providing advice until revisioe th Regulationse th f no will have taken effect. . EXPLANATOR2 Y NOT REPROCESSEN EO D URANIUM TRANSPORT
2.1. REPROCESSED URANIUM SUBJECT TO TRANSPORT procese Th s flo reprocessef wo d uraniu mshows i Figurphysicae n i th ; e1 l properties are shown hi Table I.
PROCESS FLOW REPROCESSED URANIUM SUBJEC TRANSPORO TT T
Reprocessing uranium trioxide (solid powder) or uranyl nitrate (solid or liquid solution) Conversion uranium hexafluoride (solid)
Enrichment uranium hexafluoride (solid)
Reconversion uranium dioxide (solid powder)
Fuel fabrication uranium dioxide (solid)
Reactor
Note: Because of differing process routes and locations it is conceivable that reprocessed uranium could also be transporte fore th f Um o n 3d i Or UF o 8 4. FIG. 1. The process flow of reprocessed uranium.
2.2. CHARACTERISTIC REPROCESSEF SO D URANIUM
Reprocessed uranium is obtained by recovery of the uranium from the spent fuel from commercial nuclear power stations spene Th . t fuel, whic cooles hi perioa r d fo tunef do , then undergoes chemical separation and purification called 'reprocessing'. Through reprocessing, uranium is separated from plutonium and fission products (FPs) and then used as fuel again within the same fuel cycle as unirradiated uranium, i.e.: conversion-enrichment- reconversion-fabrication-reactor. Spent fuel contains FPs, transuranic elements (TRUs) and uranium isotopes, i.e. 236232 d Uan generated only during reactor operation. AppendixI describe e radioactivitth s f naturao y d enrichean l d uranium. Appendi I giveI x a fuls l description of the nuclides that may be found in reprocessed uranium.
Throug reprocessine hth 106 gd TRUd Rprocessan uan c s suc s suc"T s FP 237,hs a h a N p and 239Pu are separated from the spent fuel. Thus the non-uranic impurity elements in reprocessed uraniu e chemicallmar y removed e compositioth t bu , f uraniuno m isotopes remains unchanged.
Although the percentage of 235U in reprocessed uranium is lower than that hi initial fuel, eithee b y r ma highe t i r loweo r r than tha naturan i t l uranium (0.71%), dependine th n go degree of fuel enrichment and burnup conditions. Reprocessed uranium usually is re-enriched to augment the enrichment of 235U to the prescribed degree. The concentrations of 232U, 234U and 236U also increase during this proces re-enrichmenf so t (whe methode nth enrichmenf so t utilizing differences in atomic mass between 235U and 238U such as gaseous diffusion and TABLE I. THE PHYSICAL PROPERTIES OF REPROCESSED URANIUM
Chemical form Physical state Soluble/ Fractional uranium content (w/w) (molecular formula) during insoluble transport
Uranium solid moderately U/U03 0,832 trioxide U03 soluble a Uranyl solid/liquid soluble U/UO2(NO3)2.nH2O 0.474 nitrate U02(N03)2
Uranium solid moderately U/UF4 0.758 tetra-fluoride soluble UF4
Uranium solid soluble U/UF6 0.676 hexafluoride UF6
Uranyl solid soluble U/U02F2 0.773 fluoride U02F2
Uranium solid insoluble U/UO2 0.881 dioxide UO2
Tri-uranium U308 solid insoluble U/U3O8 0.848 octoxide
a Generally, the water of crystallization in the solid is n = 6, and so the uranium content is 0.474 in that form. liquin I d form uraniue th , m content depend concentratione th n so .
centrifuge separatio used)e nar . Attention mus paie tb o dt U, particularl t accounti s ya r sfo larga e percentag radioactivite th f eo reprocessef yo d uranium 234.
Particular factor consideree b o st d are:
) Re-enriche(1 d reprocessed uraniu s highemha r specific activity becaus t containi e a s proportionately larger amount of U, and the uranium isotopes U and U that did not exist before irradiation. 234 232 236
(2) Trace amounts of FP and TRU can be present as radioactive impurities. dose highee Th eb rat n ) erca (3 than tha unirradiatef o t d uraniu mainle m daughtedu a o yt r nuclidf o e Th, fro e decamth f o yt higheU A . r concentrations n alsThca o contribut 228 e significantl 232 y to reducing the annua 228 l limit on intake (ALI).
) Attentio(4 n mus paie b temptie o dt d uranium hexafluoride cylinders froms whicha 6 hUF been vaporized sinc daughtee eth r nuclide Th remain 'heelse cylindee th th n n si i ' r in addition to 234Th (from the deca y228 of 238U).
Appendix III discusses particular aspects of the Regulations that may apply to reprocessed uranium. 3. DEFINITION OF UNIRRADIATED URANIUM
e definitioTh f unirradiateo n d uranium Regulationse giveth n n i t adequat no s i r efo examining conformity of reprocessed uranium to the Regulations (see Appendix IV). Both reprocessed uranium and uranium contaminated by reprocessed uranium can be classified as unirradiated uraniuRegulations.e th f o para y m9 b als s i 14 .t oI necessar mako yt cleaea r distinction for reprocessed uranium because the 1990 As Amended version of the Regulations includew no footnotn si Tablo t ) d e: ( "TheseI e applt valueno reprocesseo yo t sd d uranium".
The following definition is adopted for the purpose of this document and was submitted and approved as a change to the Regulations in the course of the continous review and revision process.
Unirradiated uranium
149. Unirradiated uranium shall mean uranium containing not more than 2 x 103 Bq of plutonium per gram of U, not more than 9 MBq (0.20 mCi) of fission products per gram 235U and not mor e235 than 5 x 10"3 grams 236U per gram 235U. existine Th g) Regulations(a restrict plutoniu masmy b s concentration (not more than 10"g 6 of plutonium per gram of U). This is unsatisfactory, insofar that it does not take accoun e widelth f o yt differing specific 235activitie f plutoniuo s m isotopes typicaA . l plutonium isotope composition arising from reprocessing operation havn sca especifia c activity 100 tunes greater than the 239P referred to in para. E-149 (SS7). The proposed definition restrict activite sth f plutoniuyo roundea mo t d down value consistent with the original definition expressed as 239P.
Adoptio f thino s value limit contributioe th s f plutoniuno m 1/100o t specifie th f 0o c activitenriche% 5 f yo d unirradiated uraniu 1/200o t e lO^Aj/ d th m f an 0 o g valur efo 239p limitatioe Th ) f fission(b o n product activity remains unchanged.
(c) The restriction in uranium-236 content is added to provide demarcation between unirradiated and irradiated uranium. Uranium-236 is formed from uranium-235 by a neutron capture mechanis presencs it d man e provide technicallsa y superior meanf so recognizing exposure to a neutron flux. The value adopted of 5 x 10" g per gram uranium-23 uppee th s 5i r consensue valuth f eo s standard ASTM-C-99 r enriche6fo d 3 uranium hexafluoride. Up to this concentration the material may be considered identical unirradiateo t d natural uranium t recognizeI . s that durin conversione gth , enrichment and fabrication processes some accidental trace contamination by irradiated uranium can occur. This specification represents the maximum value for uranium isotopes for which compositio Ae n2th consideree valub n eunlimitedca e b o dt . Note e footnotRegulationse th f Th Tabl o f : ) o d ( eeI then becomes "These values apply only to unirradiated uranium".
10 . APPLICATIO4 REGULATIONSE TH E F NO TH O T TRANSPORT OF REPROCESSED URANIUM
4.1. IKE REGULATIONS
SAGSTRAM has acknowledged that the current edition of the Regulations is unsatisfactory when applie certaio dt n transport reprocessef so d uranium. Having established a suitable definition of unirradiated uranium, it is therefore necessary to define how the Regulations should be applied to reprocessed uranium transport. This section provides a rational r thisfo e , with expansion e Regulationsth f o s where appropriat o clarift e y uncertainties or inconsistencies.
4.2. APPLICATIO REGULATIONSE TH F NO
4.2.1. Principles of the Regulations
value Th reprocesser (aAf eo )fo 2 d uranium shal determinee b l d usin compositioe gth f no uranium isotopes and the amount of other radioactive impurities contained hi the material accordancn i , e wit "formule hth mixturesr afo " stipulate5 30 d paraan n d i 4 s30 Regulations.e ofth Whereve chemicae rth l fornuclida f mo knows ei permissibls i t ni e valu2 A e e th relate knows e it us o dt o t n solubility class, takin forme gth s appropriate normao t accidend an l t conditions.
4 ) Reprocesse(b d uranium whose specific activity doe exceet sno d 10" A2/g shal classifiee lb d as LSA-II materiae th take e f b o unlimite.y s ng Aa l m contain2ma 0 1 f di s less thae nth r thafo t I mixtureAL sucd an ,h materia alsn ca ol thu classifiee sb LSA-IIs da .
) (c Reprocessed uraniu t includem no abov) (b n dei whose specific activity doe t exceesno d 3 2 x 10' A2/g shall be classified as LSA-III if it can meet standards on leachability. (d) Uranium in a powder state may be classified as LSA-III provided it can be demonstrated that there is no dispersal from a damaged package as defined hi para. 564(b)(i Regulations.e th f o )
The formula for determining the value of A2 is specified as follows in paras 304 and Regulations:e th f o 5 30
A2for mixture - S. -^ ' A2(i)
Where f(i) : fraction of activity of nuclide i hi the mixture, and A2(i): appropriat nuclide valu2 th e A r e, fo e.gei .
q B 8 10 X 232 3 U: q B 9 10 X 234 1 U: 235limio n U:t 236U: 1 X 109Bq 238U: no limit
r exampleFo Appendie sse x VIII.
11 4.2.2. Reprocessed uranium classified as LSA-II
LSA-II is defined as follows in the current Regulations:
"131. Low specific activity (LSA) material shall mean radioactive material which s bnaturlimitea yit s eha d specific activity r radioactivo , e materia r whicfo l h limit f estimateo s d average specific activity apply. External shielding materials surroundin materiaA LS e gconsiderele th shalb t lno determininn di estimatee gth d average specific activity.
LSA material shall be in one of three groups:
) (a LSA-I (not considered here).
(b) LSA-II
(i) Water with tritium concentration up to 0.8 TBq/L (20 Ci/L); or (ii) Other materia whicn i l activite hth distributes yi d throughoue th d an t estimated average specific activity doe exceet sno d 10^ A 2solidr /gfo s 5 and gases 10"d an , A r 2liquids/gfo . ) (c LSA-III (not considered here)".
Reprocessed uranium in the solid state which meets either of the following criteria shall classifiee b LSA-IIs da :
4 ) (1 Reprocessed uranium material whose specific activity doe t exceesno d 10" A2/g. (2) Reprocessed uranium material 10 mg of which does not exceed 1 ALI taking into consideration its chemical form and the appropriate inhalation classes D, W or Y of ICRP publication botr fo h] 5 norma , acciden[4 sd an l t situations.
According to ICRP 30 [5], uranium compounds whose inhalation class is D (water soluble) are UF6, UO2F2 and uranyl nitrate; those whose inhalation class is W (moderately soluble) are UO3 and UF4; UO2 is inhalation class Y. Where the likely degree of solubility or chemical form (organic or inorganic) of a particular radionuclide s knowi e inhalationth nchemicae clasth f o s l form under both norma accidend an l t conditions shoul takee db n into account usin mose gth t appropriate valu currentls ea y recommended by ICRP. A fuller discussion of the application of LSA-II to reprocessed uraniu foune b Appendin n di m ca . xV
4.2.3. Reprocessed uranium classified as LSA-III
Reprocessed uranium whose specific activity exceeds lO^A^g (the upper limit for LSA-II) e sam, th whict ea tims i hn insolubl a e e compound whose specific activits i y 3 2 x 10~ A2/g or less, shall be deemed as LSA-III if it meets the requirements specified in para. 131 of the Regulations:
" 131. Low specific activity (LSA) material shall mean radioactive material which by its nature has a limited specific activity, or radioactive material for which limit f estimateo s d average specific activity apply. External shielding materials
12 surroundin materiaA S L e consideregle th shalb t no l determininn di estimatee gth d average specific activity.
LSA material shall be in one of three groups:
(a) LSA-I (not considered here).
(b) LSA-II (not considered here).
(c) LSA-III
Solids (e.g. consolidated wastes, activated materials) in which:
(i) The radioactive material is distributed throughout a solid or a collection of solid objects, or is essentially uniformly distributed in a solid compact binding agent (suc concretes ha , bitumen, ceramic, etc.); (ii) The radioactive material is relatively insoluble or it is intrinsically containe relativela n di y insoluble matrix, thato s , even under losf so packaging e los f radioactivth ,so e materia r packagpe l leachiny eb g when placed hi water for 7 days would not exceed 0.1 A2; and (iii) The estimated average specific activity of the solid, excluding any 3 shielding material, doe exceet sno x 10" d2 A2/g." 4.2.3.1. Uranium dioxide ceramic pellet LSA-IIIas material
relevane Th t requirement r LSA-IIsfo summarizee ar I followss da :
(1) It must be a solid in which radioactive material is uniformly distributed.
3 (2) Its specified activity does not exceed 2 x 10" A2/g.
(3) Radioactive material exceeding 0.1 A2 does not leach from a package during a leaching tes waten days7 i t r fo r.
The leaching tests for 7 days on uranium dioxide pellets conducted hi the UK and Japan 199n i 0 (Appendi ) confirmexVI d tha valua t e adequately lower tha s standare i nth 2 A 1 d0. obtained with this material.
Based on the definition of LSA and the results of these leaching tests, uranium dioxide pellets (including the case of a fuel assembly) satisfying these criteria can be deemed as LSA-III.
4.2.3.2. Uranium dioxide triuraniumand octoxide powders LSA-IIIas material
leachine Ud Th 3Oan 82 gpowde testUO Japad n so an r 199n nconductei K U 0 e th n di (Appendix VI) also confirmed that these materials satisfy the standard leaching test of less 2 valuA tha 1 e n0. afte r immersio watedaysn n7 i r fo r.
However, sinc e referencth e LSA-IIo t ecurrene th n i I t Regulations refer nono t s - dispersible radioactive materials, the additional requirement that there is no dispersal from a damaged package under conditions likely to be encountered during normal conditions of
13 n accidena transpor n i d t an shalt e applieb la packag o t d e intende o transport d t such reprocessed uranium oxide powder (Appendix VII) e criteri.Th r establishinfo a g this requiremen specifiee ar t paran di . 564(b)(i Regulations.e th f o )
4.2.4. Consideratio daughtee th f no r nuclide uraniuf so m
According to para. 303 of the Regulations, when calculating the values of Al and A2, whe ndecaa y chai i whicnh hdaughtea r nuclid half-lifa s eha e longer either tha dayn nte s or that of the parent nuclide, such parent nuclide and daughter nuclide are to be considered a mixtursa f differeneo t nuclides calculatinn I . value gth f Af reprocesseeo 2o d uraniumt i , is therefore necessar o takt y e 228Th (daughter nuclid f 232o e U) into account (see Appendix VIII).
Among the daughter nuclides of reprocessed uranium 228Th should always be taken into account. While undisturbed the amount of 228Th builds up to reach the maximum value in 228 approximately 10 years. When calculating the A2 value it should be assumed that Th is in full equilibrium with 232U to give the maximum figure.
4.2.5. Classificatio f enricheno d commercial grade uranium
The new definition of unirradiated uranium, which corresponds to enriched commercial grade uranium (ECGU t ASTMa ) havn ca ,specifiea c activity very slightly greater than 4 10"enrichmento w/ A5 2t /ga . However, thi alss i s o tru f completeleo y naturally derived enriched uranium under some enrichment conditions, although thi s alwayi s s specifies a d 'unlimited A2'. It is concluded therefore that ECGU differs insignificantly from this and may also be accepted as 'unlimited A2' within the specified composition. (Appendix IX details the calculations).
4.2.6. Observations on radioactive impurities
Although trace quantitie f radioactivo s e impuritie e containear s i reprocesseh d d uranium, their contribution to the A2 value may be neglected in principle. In calculating the A2 valu reprocessef eo d uraniu accordancmn i Regulations,e th f o 5 e wit30 d h an para 4 s30 sufficiens i t i tako t t e uranium isotope theid san r daughter nuclide wels s a representativ s a l e nuclide radioactive th f so e impurities (FP, TRU) into account.
In calculating the A2 value of reprocessed uranium to classify it into one of the LSA categories importans i t i , t alway tako st e daughter nuclide uraniuf so m isotope nuclided san s of radioactive impurities (FP, TRU) into account. It shall be confirmed whether nuclides of radioactive impurities (FP , takiny neglectee TRUb b t n gno )ca their do r representative nuclides into calculating the A2 value. (Refer to Appendix X). 4.2.7. Consideration of changes in dose equivalent rate with time
Since the increment of gamma ray dose-equivalent rate by 208T1 increases hi proportion build-ue toth 228f po Th, whic firse th t s hi daughte r nuclid 232 f importanes o i U t i , controo t l build-ue th p period fro stage mth whict ea h daughter nuclide chemicalle sar y separated from reprocessed uranium (for instance fro poine mth t whica t hcylindea filles i r d with UF6). The dose equivalent rate of a package containing reprocessed uranium is higher than one containing unirradiated uranium due to differences in radiation source specifications. This
14 is primarily becausT1e,208 which emits high energy gamma rays, is produced in the decay chain of 232U which is not found in unirradiated uranium.
208T1, which rapidly reaches radioactive equilibrium with 228Th (half-life: 1.91 years), firse th t daughter nuclid 232f eo U, show maximusa m activity valu aboun ei yearsn te t . (See Appendix XI).
Accordin resulte Germae th th o gf t so ncylindersY stud48 f yo , 'Investigatione th n so Transport of UF6 in 48Y containers with Uranium from Reprocessed PWR Fuel Elements' (Appendix XII), the dose equivalent rate of a cylinder filled with uranium hexafluoride, peaa build o approximateln t ki p su yearsthen s i y te t n bu stil, l adequately belo surface wth e dose equivalent rate limits stipulate Regulationse th n di mSv/h)2 ( . After emptyin UFe gth 6, the 228Th remains in the heels within the cylinder but is no longer shielded by the bulk of the uranium. Such a cylinder can then exceed the maximum allowable surface dose rate for transport, depending on the initial composition of the uranium and its storage tune while full.
4.2.8. Handling of cylinders emptied of uranium hexafluoride Whe a ncylinde r filled with reprocessed uranium hexafluorid e re-usedb o t s ,i e consideration must be given to removing heels by cleaning the cylinder before transport of the empty cylinder.
Taking the example of an empty 48Y type cylinder which had been filled with reprocessed uranium before enrichment, the heel specific activity exceeds the limit for LSA- II, even if the maximum permissible heel amount (22.68 kg) is assumed. This is primarily contributioe duth o et 228y nb Th . Moreover, uranium hexafluoride cannot mee insolubilite tth y criterio LSA-IIIr nfo addition I . totas nit l activity exceed mixture value th s th r Af ed o 2fo ean hence it cannot be classified as Type A (Appendix XII). The contribution of 228Th becomes much greater after enrichment as a result of the enrichment of U. 232 It should be noted that, if heels are not removed, the contamination level inside the cylinder also exceeds the limits for an empty container, and for SCO-I, and SCO-II.
4.2.9. Additional requiremen typP I en a packag r fo t e containing reprocessed uranium General testing conditions for IP-2 or IP-3 packages specified in paras 519 and 520 of the Regulations to be replaced for transport of LSA-III powders by those prescribed by para. 564b(i) (tests for demonstrating ability to withstand normal and accidental conditions of transport). 4.2.10. Approval by the authorities and multilateral approval of reprocessed uranium transport
As indicated in Sections 4.2.1-4.2.8, the transport of reprocessed uranium, which takes accoun f chemicao t l form shal e subjecb l o approvat t l (design y nationab ) l authorities; international transport shal subjece b l multilaterao t t l approval.
4.3. MARKING OF PACKAGES AND INSPECTION BEFORE SHIPMENT Marking, labelling and placarding concerning transport shall be in compliance with requirements specified in paras 436-445 of the Regulations. It is not necessary to give a special indication of 'reprocessed uranium' during transport.
15 Package inspection to the requirements specified in para. 402 of the Regulations should e madb e before each shipmen confiro t t e compliancmth f transporteo e d item theso t s e requirements. Special attention shoul e pai b do confir t d e amounmth f radioactivito t y contained in the package, if reprocessed uranium is to be transported. It should be confirmed thae reprocesseth t d uranium containe e packaginth n n i complianci d s i g e wite th h requirement radioactivitn so y limits prescribe LSA-IIr dfo r LSA-IIIo I materials (para1 31 . of the Regulations).
e overalTh l compositio f reprocesseno d uranium depend e typ th f reacto o en o s n i r which the original uranium has been irradiated, the burnup, the cooling history after irradiation decontaminatioe th , n efficienc subsequenf o y t purification processe time th ed san thaelapses ha t d after reprocessing.
Consequently package th n i , e inspection conducted before each shipmen mades i t e th , r reprocessevalufo 2 A e d uranium mus e calculate b te percentag e basith th f o sn do f o e uranium isotopes and radioactive impurities (FP/TRU, etc.) in the reprocessed uranium, using the "formula for determining the value of A2 for a mixture" specified in the para. 304 of the Regulations.
After considerin chemicas git l form under norma accidend an l t conditions, reprocessed 4 5 uranium material whose specific activity doe t exceesno d 10" A2/g (10~ cas e Ath f 2 eo /n gi liquid classifies i ) d into LSA-II reprocesse.e Alternativelyth f o g m d0 1 uraniu f i , m doet sno exceeI (annuaAL d1 l limi intakealsn y o classifiete ob ma t i ) LSA-IIs da .
Solid reprocessed uranium material whose specific activity exceed above th s e value (10" A t 2doe /gt exceebu ) no sx e 10"classifie b 2 dy A 2/s LSA-IIIma ga d . Powdered reprocesse4 d uranium3 may be classified as LSA-III if the package meets the tests specified in para. 564(b)(i) without dispersa contentse th f o l .
package th n I e inspection befor actuae eth l shipment compliance th , contente th f eo f so the reprocessed uranium material shall be confirmed by analysis and evaluation.
16 5. SUMMARY
The consultants services meetings have examined the transport of reprocessed uranium e variouth n i s IAEA Safety Series document havd an se establishe dmethodologa y which allows full compliance intRegulations.e oth categorpracticA n I LS e yeth wil applicable b l e o most t such transport movements. Some clarificatio e Regulationsth f o n s needei o t d demonstrate this.
existine Th g definitio f unirradiateno d uraniu minadequates i , becaus t woulei d allow reprocessed uraniu classifiee b mo t unirradiateds da . Consequently bees ha nt i ,propose o dt chang definitioe eth unirradiatef no d uranium encompassuco t s ha s uranium containint gno more than 2 x 103 Bq Pu/g235U of Pu, not more than 9 MBq/gU of fission products and not mor '30 g1 e 236 thax U/ n5 g 23SU.
The classification methods for reprocessed uranium material for transport are summarized as follows:
(1) Reprocessed uranium material in which the activity is distributed uniformly and which fulfills either of the following requirements l(a) or l(b) shall be classified as LSA-II.
e specifiTh ) c (a activit f solio y d reprocessed uranium material doe t exceeno s d 4 10" A2/g where the value of A2 is calculated for all significant nuclides in the material. For liquid reprocessed uranium material the specific activity of the 5 material shal t exceeno l d 10" A2/g.
TABLCLASSIFICATIOA LS . EII MAIE TH N F NREPROCESSEO D URANIUM MATERIALS
Form Activity content Classification Comments
Uranium trioxide less than lO^/g LSA-II
5 Uranyl nitrate less than K)- A2/g LSA-II Relevant chemical forms are: UO2(NO3)2, U308* Uranium less than lO^/g LSA-II Relevant chemical hexafluoride forms are: a UF4, U02F2 Uranium less than lO^/g LSA-II 3 b dioxide powder less thalQ-X n2 A 2/g LSA-III Uranium dioxide pellet OR less than lO^Aj/g LSA-II fued ro l OR 3 assembl y/ bundl e less than 2 X 10' A2/g LSA-III
3 The chemical forms indicated are all those considered by the consultants as being likely to arise in significant quantities from a transport accident involving uranyl nitrat r uraniueo m hexafluoride. b Package must not allow dispersal under normal and accident conditions during transport.
17 (b) When 10 mg of reprocessed uranium material contains in radioactivity less than takinI AL g1 accoun compouns it f o t d for solubilitd man y class under normad lan accident conditions.
(2) Solid reprocessed uranium material whose specific activity exceeds lO^Aj/g (the upper 3 limi r LSA-II)fo t t doe t exceebu , no sx 10~ d2 Awhicd 2/gan h meet leachine th s g standard as defined in para. 131(c)(ii) of the Regulations shall be classified as LSA-III. Uranium material in the powder state must meet the additional performance requirements tha materiao n t disperses i l d fro package mth e under condition f tesso t specifie paran di . 564(b)(i Regulations.e th f o ) classificatioA LS e maie Th th nf n o reprocesse d uranium material shows si Tabln i . eII
18 AppendixI THE RADIOACTIVITY OF URANIUM
The total radioactivity of uranium at a given enrichment depends on a number of factors including the origin of feed material, the method of enrichment, the period of time after processin uraniu e percentage th th f go d man f otheeo r radioactive impurities.
relatioe Th n between specific activit enrichmend yan uranium-23f o t 5 (SS37, 1990s i ) show n Tabli n e III e valu. Th f specifio e c activit f uraniuo y m include e activitth s f o y uranium-234 enriched during the enrichment process for uranium-235 but does not include contribution uraniuy b s m daughter nuclides r uraniue valuefo Th e . ar sm enrichee th y db gaseous diffusion method starting from natural uranium. If the origin of source material is unknown, the specifis nit c radioactivity mus measuree b t r calculatedo basie th datf sn o do a on isotopic ratios.
Since uranium fuel cycle compound producee sar d through chemical processes, their equilibrium with uranium daughter nuclides may be disturbed during those processes. Whereas uranium isotopes have long half-lives daughtee th , r nuclide r uranium-23so d 8an uranium-235 (thorium-23 thorium-231d 4an , respectively) have short half-lives. Therefore equilibrium is almost reached in approximately 150 days. After this period, for instance, 4 accordin IAEA-TECDOC-608go t specifie th , c activitreache6 UF Bq/ f yo s10 g nearlx 6 y3. at 0.45% enrichment or 4.0 x 104 Bq/g at natural enrichment.
The enrichment process also enriches the minor isotopes of uranium (i.e. U, U and 236U). As the ratio of 234U to 235U is greater in reprocessed uranium than for naturall 232 y 234derived uranium at the same U enrichment, this isotope contributes significantly to the increased 235 specific activity of enriched reprocessed uranium.
TABLE III. SPECIFIC ACTIVITY VALUES FOR URANIUM AT VARIOUS LEVELS OF ENRICHMENT
Mass percenf o t Specific activitya uranium-235 presenn i t uranium mixture Bq/g g/Bq 0.45 1.8 X I04 5.6 X IQ'5 0.72 (natural) 4 10 X 6 2. 3.8 x 10'5 1.0 4 10 x 8 2. 3.6 X IQ'5 1.5 4 10 3. 7x 2.7 x IQ'5 5.0 1.0 x 10s x lO' 0 5 1. 10.0 1.8 x 10s 5.6 x IQ'6 20.0 3.7 x 105 IQ'X 67 2. 35.0 7.4 x 10s 1.4 X 10-6 50.0 5 10 9. 3x IG'X 61 1. 90.0 2.2 x 106 10-X 75 4. 93.0 2.6 x 106 3.8 x IQ'7 95.0 3.4 x 106 2.9 x IG'7
These valuespecifie th f so c activity includ activite eth uranium-23f yo 4 whic concentrates hi d durine gth enrichment process, and do not include any daughter product contribution. The values are for the material originating from natural uranium enriched by a gaseous diffusion method. If the origin of the material is not known the specific activity should be either measured or calculated using isotopic ratio data.
Next page(s) left blank 19 Appendix II NUCLIDES COMPOSING REPROCESSED URANIUM
Uranium compounds such as UF6 and UO2 produced from unirradiated uranium do not include sources of radioactivity other than from uranium isotopes existing in nature, i.eU.,238 235U, 234U and their daughter nuclides.
Uranium material UOd an 2 6 sgenerate sucUF s ha d from uranium separated from spent fuel through reprocessing (reprocessed uranium) can contain the following nuclides in varying concentrations:
) Uraniu(1 m isotopes r instancefo , , U, U', U', U, d UUan . 232 234 235 236 237238 J (2) Transuranic isotopes (TRUs), for instance, 237Np, 239Pu.
(3) Fission products (FPs) instancer fo , , 106Ru, 95Zr/95 "Tcd Nban .
(4) Daughter products of these nuclides, for instance, 233Pa and 228Th and 208T1.
compositioe Th f nuclideno s containe reprocessen di d uranium depend type th f eo n so reactor in which irradiation has taken place, its burnup and subsequent cooling history, impurity decontamination efficiency Preprocessing, enrichment, and reconversion processes, and time that has elapsed after such processes have been completed. Processes which change radioactive equilibria, e.g. solvent extractio r vaporizationo UFf n o important e 6ar , .
Under these conditions, indices for transportation in accordance with the Regulations mus calculatee b t eacr dfo h transporte dbasie ite th f specifi so n m o c analysis data.
specifie Th c activit othed yan r propertie f variouso s type f reprocesseso d uraniue mar show i Tablnh . eIV
1 Uranium isotopes existin naturen gi .
21 to TABLE IV. CHARACTERISTICS OF REPROCESSED URANIUM PACKAGE
Enriched uranium of natural source ECGU' ERUa Remarks
Item U UF6 U02 UF6 U02 Enrichment 5% 5% 5% 5% 5%
9 9 8 8 A2(Bq) Unlimited 1.1 x 10 1.1 x 10 8.8 x 10 8.8 X 10
1 - 4 £ /(0 ' ^2 (0
Quantity equivalent of A2 (kg) Unlimited 12.2 9.2 2.3 1.8 Activity (Bq/g) 10X 5 0 1. 4 10 X 0 9. 1.2 X 105 3.9 x 105 5.0 X 10s Activity limit of LSA-II 1.1 x 10s 1.1 x 10s 1.1 x 105 8.8 X 104 " 10 x 8 8. f(i): fractio f activitno f yo (10%/B) nuclide i
Activity limit of LSA-III A2: appropriate A2 value 3 s 6 6 (2 X 10' A2/g) 2.0 x 10 not applicable 10 2. 2x not applicable 1.8 X 10 for the nuclide i
ECGU: enriched commercial grade uranium (ASTM C-996) maximum values. 1 ERU: enriched reprocessed uranium (ASTM C-996) maximum values. Appendix in IAEA REGULATIONS CONCERNING REPROCESSED URANIUM
This appendix summarizes those standards specifie IAEe th i Ah d Regulatione th r sfo Safe Transport of Radioactive Material (the Regulations) that are especially pertinent to this document IAEe Th . A periodically revise Regulations.e sth
Regulationse Th have serve modea regulatora s da r fo l y rule establishe Membey db r Statepublishea r wels fo sa s a l d rule institute internationay db l transport organizationse Th . latest revisio Regulationse th f no publishes wa 1985n di . Supplements were issue 198n di 6 and 1988. As amended versions of the Regulations and its supporting documents SS7, SS37 and SS80 [6] were published in 1990.
Belo summarizee war majoe dth r requirement Regulationse th n si concerning UF6, UO2 powder and pellets, fuel rods, and other reprocessed uranium materials described in this document.
Regulationse Th consis followine th f o t g requirement contentn so packagingd san :
(1) The quantity of LSA material or SCO in a single industrial package Type I (IP-I), industrial package Type 2 (IP-2), industrial package Type 3 (IP-3) or object or collection of objects, if appropriate, shall be so restricted that the external radiation fro m unshieldee levem3 th t a l d materia r objeco l r collectioo t f objectno s doet sno exceed 10 mSv/h (1 rem/h) (para. 422).
(2) The radioactive material is distributed throughout the nuclear fuel or in a matrix, (para. 131(b)(ii) and (c)(i)).
e specifiTh ) c (3 activit f reprocesseyo d uranium shal t exceeno l e standardth d valus a e follows:
(i) LSA-II:
- For distributed solids: "The estimated average specific activity does not exceed 4 10- A2/g" (para. 131(b)(ii)).
5 liquidsr Fo :- "The estimated average specific activity dos exceet eno d 10" A2/g" (para. 131(b)(ii)).
(ii) LSA-III:
"The estimated specific activit solide th f y,o excludin shieldiny gan g material, shall 3 not excee x 10' d2 A2/g" (para. 131(c)(iii)). e cas th f LSA-III o en I ) ,(4 evea containee radioactiv f th i n f i s d losi r an te material transported is left in water for 7 days, the amount of radioactive material leaking to water per transport unit shall not exceed l/10th of the value of A2 (paras 131(c)(ii) and 603).
(5) A transport containing reprocessed uranium which is both fissile and classified as LSA II/LSA III shall be subject to:
23 package th - e requirement (para O materiaA SC . LS d 426 r an lsfo , Tabl; eV)
- the requirements for containment and shielding for IP-2/IP-3 packages (paras 518-523);
requiremente th - fissilr sfo e material (paras 559-568).
(6) Total radioactivity of one transported item of LSA material shall be limited so as not to exceed the radiation level stipulated in para. 422.
additionn I totae th , l activit r conveyancype limites ei d (para. 427, Table VI).
(7) Transported matter in excepted packages, including transported matter containing fissile materiaemptn a t d exceedinyan no l6 cylindeshal, g UF 5 lr 1 gmee fo re sam th t e requirements as natural uranium for transport and control (paras 415-421).
24 AppendiV xI DEFINITIONS OF UNIRRADIATED/REPROCESSED URANIUM AND THEIR IMPLICATIONS
INTRODUCTION
In the current Regulations, the definition of unirradiated uranium defines limits for fission products and plutonium, for the purpose of excluding any uranium which has been irradiated in a nuclear reactor. But because of technical improvements in the efficiency of reprocessing purification processes, this limitatio n fissioo n n produc d plutoniuan t m concentratio longeo n s ni r effective.
Table I of the Regulations (as amended 1990) in providing A/A2 values adds a footnote e 'unlimitedth o t ' A2 values ascribe (natural)U o t d (enricheU U ,r lesso d an )% 5 d (depleted excludo t ) e reprocessed uranium reprocesses A . d uraniu t definee no th m s i n i d Regulations, implicatioe th thas ni t uranium which doe t mee definitiose no th t n unirradiated must be considered to be 'reprocessed'.
After extensively having studied alternative experte sth s concluded tha cleata r definition of unirradiated uranium would remove any need to define a new category 'reprocessed uranium'.
e reasoTh r defininfo n g 'unirradiated uraniumy virtub e exclusiod th e an ' f o n 'irradiated' (reprocessed, recycled) uranium is to ensure that the special characteristics of irradiated uranium are recognized in assessing safety and the technical suitability of recommended practices for transport.
Further consideratio definitiow f suco nne ha f unirradiateo n d uranium raisee dth practical issue as to whether the commercial category of uranium, ECGU (ASTM specification), coul acceptee db s 'unirradiated'da . This topi deals ci t wit exampley hb n si calculations presente attachmentse th n di .
DEFINITION OF UNIRRADIATED URANIUM
In terms of the 'definition of unirradiated uranium' in para. 149 of the Regulations, an investigatio mads nwa e into criteri discriminatinr afo g between unirradiate reprocessed dan d uranium.
(1) Regulations (1985 Edition)P F , Pu :
This definitio removee b nn alonca d P unsatisfactors ei F fro d man irradiateu P s a y d uraniu solvenmy b t extractio other no r chemical processing technolog valueo yt s below the limits of specified. This improvement in reprocessing technology could therefore allow some uranium to be designated as 'unirradiated' even after exposure in a reactor.
25 (2) UK proposal: Pu, FP, 236U
Whilst chemical purification technique removn sca e associated radioactive species such as fission products and transuranic elements, the purified uranium, because of its irradiation, will contain the minor isotopes uranium-232 and uranium-236. These minor isotopes are technically significant to subsequent safety considerations and fuel propertie alsd san o provid meanea f identifyinso g prior irradiation history. Because uranium-23 produces 6i greaten di r abundanc becausd ean e measurement techniquee sar simpler for this isotope, it is the preferred identifier for irradiation.
) Investigation(3 ) (a U , Japan:n FP i , Pu
Whilst measuremen U(af o t ) activit relativels yi y direca simpls i d t measurean e th f eo principal activity associated with uranic materials, it is significantly affected by the concentratio naturalle th f no y occurring isotope uranium-234 ratie uranium-23f Th o.o 4 totao t l uranium falls with irradiatio lessea o t t r degrenbu e than uranium-235 during reactor burnup. This means only after re-enrichment is the ratio of uranium-234/uranium-235 enhanced relativ naturao et l product henc — shighee eth r specific activity.
The measurement of U (a) is therefore not a satisfactory discriminator for irradiation in all circumstances. Variations in enrichment processes, e.g. gaseous diffusion or centrifuge, type of feed, and operating tails assay, can also modify the ratio uranium- 234/uranium-235. This has implications for the values for specific activity ascribed to enriched (natural origin) materials.
Therefor followine eth g definitions relatin unirradiateo gt d uranium shal addede b l :
Unirradiated uranium
The Regulations:
149. Unirradiated uranium shall mean uranium containinf o q B 3 t morg10 no e X tha n2 plutoniu r gram pe uranium-235f mo t mor no ,(0.2 q e tha0MB nmCi9 f fissioo ) n products r grape m uranium-23 t morno ed 5tha10~an X nurôniurn-235 g r gra6pe m uranium-235. 3
Footnote d Table I
These values apply onl unirradiateo yt d uranium.
557
E-149 tere mTh . 'unirradiated uranium intendes i ' excludo dt uraniuy ean m whic bees hha n exposed to a neutron flux so as to transform some of the uranium-238 into plutonium-239 and some of the uranium-235 into fission products. The limits of 2 x 10 Bq plutonium per gram of uranium-235 and 9 MBq (0.20 mCi) of fission products per gram3 uranium-235 are intended to identify the presence of irradiated uranium whilst recognizing the presence of trace amounts of plutonium and fission products in all natural uranium.
26
presence Th f uranium-23eo mora s 6I e satisfactory indicatio f exposurno neutroa o et n X 10'flux gram5 3 . s uranium-23 r grape 6m uranhim-23 s beeha 5 n choses a n representing the consensus view of ASTM Committee C-26 in specification €-996 for enriched commercial grade uranium. This value recognize e possibilitth s r tracfo y e contamination by irradiated uranium but provides that the material may still be treated as unirradiated. This specification represent maximue sth m valu r uraniufo e m isotoper fo s which composition the A2 value can be demonstrated to be unlimited for uranium hexafluoride. The difference in À2 for uranium dioxide is considered to be insignificant. e define Th 0 hav15 de d termbeean paran i n9 s combine14 s paran i d . 131(a)(iio t ) prescribe acceptable forms of LSA-L
Note: Changes from the 1985 version (as amended 1990) of the Safety Series documents shadede ar .
27 Attachmen Appendio t 1 t V xI DISCUSSIO DEFINITIOF NO UNIRRADIATEF NO DM URANIUCS T 1S T MA definitioe Th ) Regulationsne (1 th give f o para t technicalln ni 9 no 14 .s i y correct because it does not formally exclude all irradiated uranium.
presene (2Th ) t definition base plutoniun do fissiod man n product contentst coulme e db by irradiated uranium after suitable reprocessing uraniu y conditionan r mfo d isotopsan e composition. presence Th ) f uranium-23eo (3 mors 6i e satisfactory evidenc exposurr efo neutroa o et n t consisteni flud xan t wit intentioe hth f excludinno g irradiated material accordino gt E-14 557f 9o . proposet no s i t I removo dt ) presene (4 eth t restriction plutoniun so fissiod man n products becaus reasone th f eo s provide paran di . E-12 f SS7.9o
(5) In order to improve the present definition, it is proposed to add the additional criterion that the material should not contain more than 5 x 10~ grams uranium-236 per gram uranium-2352. 3
(6) This value has been chosen as representing a consensus view of ASTM Committee C-2 specification 6i n C-996. This value will limit plutoniu fissiod man n products more stringently thae existinth n e readilb g n definitioca y d demonstratean n y b d measurement.
e evaluatioTh f uranium-23no provisionas 6i wild an ll require further evaluation. 2
28 Attachment 2 to Appendix IV PROBLEM UNIRRADIATED/REPROCESSEF SO D URANIUM FUEL CYCLES
e definitionTh f unirradiateo s d reprocessean d d uranium must mak a distinctioe n between four types of uranium, as indicated in Figure 2.
Unirradiated uranium cycle
Natural uranium Enriched uranium
1. Enrichment Unenriched ^^^^^^^ Enriched
Contaminatio. 4 n 3. Contamination
Reprocessed uranium cycle I 5.1 Mixin^ g
Reprocessed 2. Enrichment Reprocessed
Unenriched ^^^^^^r^ Re-enriched
FIG. Interaction2. zones of unirradiated/reprocessed uranium.
(a) If unirradiated and reprocessed uranium are treated at the same time in enrichment and fabrication facilities, then contaminatio unirradiatef no d uraniu reprocessemy b d uraniuy mma occur. Therefor necessars i t ei examino yt permissible eth e exten thif o t s contaminatiod nan establish such limit accordancn si esyste Q wit e hf IAEAmth o , taking into account ASTM and other commercial standards. (ProbleFigurf o 5 . d e2) man arisin4 , 3 n gi
(b) It is unreasonable that what is classified as unirradiated or natural uranium could be classified otherwise after an enrichment process. In order to avoid this possibility, and taking into account existing enrichment process technology, an internationally accepted composition of "commercial natural uranium" (CNU bees ha )n produce ASTy db M Committee C-2n i 6 the form of a specification for the feed to an enrichment process (ASTM C-787). Within this specification the maximum trace contamination by irradiated uranium is controlled by limiting 'irradiatede th ' uranium isotopes:
uranium-23 t greate2no r than 0.0 x 110' 9g/gU
uranium-236 not greater than 20 x 10"6g/g U
On enrichmen uranium-23% 5 o t t 5 this material will satisfy C-996 specificatior nfo "enriched commercial grade uranium" (ECGU).
The justification is that CNU and ECGU can be treated in all respects as being the same as uranium of pure natural origin.
29 purpose definitioe Th th f eo f unirradiateo n d uraniu m"excludo t e uranium usea n di reactor" may remain unaltered, but it must recognize the possibility of contamination of unirradiated uraniu reprocessemy b d uraniu lighe th parallef o unirradiatef mtn i o e lus d dan reprocessed uranium.
(c) The specific activity of ECGU to the ASTM C-996-90 standard (upper limits for specifications of 5.0% enrichment) and the domain where "10 mg is 1 ALI or less" are shown in the Figure 3, along with specific activity in relation to enrichment indicated in the 1973 Regulations. The specific activity of ECGU to the ASTM Standard in the form of uranium hexafluoride is below the specific activity of LSA-II, but it can, in the form of uranium dioxide, conceivably exceed the specific activity limits for LSA-II.
However, the specific activity of 5% enriched uranium arising from purely uncontaminated natural source dominates si concentratioe th y db uranium-23e th f no 4 isotope. e currenTh t Regulations t mandate variabilitno th o r d fo e f uranium-23o y enrichen 4i d products arising from commercial enrichment processes. The ratio 10 000 jig uranium- 234/gram uranium-23 t unusuano s 5i enrichmenn i l t plant tolerances (ASTM C-996t a d )an this level the effect of the minor isotopes uranium-232 and uranium-236 as specified for ECGU is insignificant. It follows that if this unirradiated uranium to 5% uranium-235 enrichment can be classed as ' A2 unlimited' (Table I of the Regulations) then ECGU at the same enrichment must als 'unlimitee ob d A2' (see Tabl. eX)
(d) As noted in (c), the specific activity of unirradiated enriched uranium is governed by the uranium-234 isotope. During irradiation as fuel in a reactor the concentrations of the uranium-235 and uranium-234 isotopes fall. It was a demonstrable conclusion of the first CSM hi 1989 that the net specific activity of uranium decreases during irradiation with loss of uranium-234 and uranium-235 even through uranium-232 and uranium-236 increase. If the activities of daughter products and transuranic nuclides are sufficiently low after reprocessing, it is possible for reprocessed uranium (prior to further enrichment) to have a
3e+05
5.0% estimated maximum samplf eo spécifie activity reprocessed 2e+05 uranium
specific activity of IAEA
1e+05-
upper limi ECGf to U (enrichment 5.0%)
10 enrichment (%)
FIG. Enrichment3. specificand activity.
30 lower specific activity tha startine nth g fuel, thus placin materiae g th same th en i l 'unlimited A2' category for transport. Irradiated uranium after reprocessing but prior to further enrichment can therefore meet the existing requirements for LSA-II without any need to chang Regulations.e eth
(e) Following further enrichment the ratio of uranium-234/uranium-235 is greater for irradiated uranium than for unirradiated. This relationship arises from two factors. Firstly, the loss of uranium-234 during irradiation occurs proportionately more slowly than fission of uranium-235 (burnup). Henc ratie eth o uranium-234/uranium-23 increaseds 5i . Secondly, during enrichment (gaseous diffusion, centrifuge) the physical processes enhance the relative concentration uranium-234/uranium-238 more rapidly than uranium-235/uranium-238, increasing the ratio uranium-234/uranium-235 in the product. As noted on previous occasions this increase specifie sth c activit uraniue th f yo m becaus highee th f eo r specific activitf yo uranium-234 isotope. Enrichment can also lead to increased concentrations of residual radioactive impurities suc fissios ha n products (106Ru, "Tc transuranid )an c nuclides (237Np, 239Pu, etc.).
Next page(s) left blank 31 Appendix V REPROCESSED URANIUM CLASSIFIED AS LSA-II
Reprocessed uranium can be classified as LSA-II in the following cases:
4 ) Wher(1 specifie eth c activit reprocessef yo d uranium material doe t exceesno d 10" A2/g. (2) Where, taking into consideration its compound form, 10 mg of reprocessed uranium material doe t exceeundeI sno AL rd1 norma accidend an l t conditions.
This criterion is based on SS7 (1985 Edition), Second Edition. The following descriptio founs ni Section di n AI.5.concernin7 55 f 5o case gth e whers i 2 value A eth f eo judge 'unlimitede b system Q o d t e th .n i '
"AI.5.5. Low specific activity materials
The 1973 Edition of the Regulations recognized a category of materials whose specific activities are so low that it is inconceivable that an intake could occur which would give rise to a significant radiation hazard, namely low specific activity (LSA) materials. These were define termn i dmodea f so l whers wa t ei assumed that it is most unlikely that a person would remain in a dusty atmosphere long enoug inhalho t e mor materialf o eg tham 0 n.1 Under these conditionse th f i , specific activit materiae th f yo s suci l h thamase th t s intak equivalens ei e th o t t activity intake assume occuo persoa dt r fo r n involve accidenn a n i d t witha Typ packageeA , namely 10" A2, then this material shoul t presendno greatea t r hazard during transport 6than the quantities of radioactivity in Type A packages. This hypothetical model is retained within the Q system and leads to an LSA limit 1 4 c og"fQ value10";Q thu x r e thos th fo s e radionuclides whose specific activity is below this level are listed as unlimited. In the cases where this criterion is satisfied the radioactivity associated with 10 mg of the nuclide is less than the appropriate ALI value recommended by the ICRP. "
That is, where radioactivity from 10 mg of the material in question does not exceed the annual limit on intake (ALI) in ICRP 30 then the value of A2 is judged to be 'unlimited' in th systemeQ . Consequently , where <1 f i , :
Radioactivity 10 mg of uranium isotope i (Bq) ALI of uranium isotope i (Bq)
then the value of A2 can be judged to be 'unlimited'.
Referenc e chemicath o t e l systeforQ e calculation mi th s m i n e valui th 2 f A eno describe follows da 557n i s .
33 "AI.4.3 Interna— c Q . l dos inhalatioa evi n
(HgU A ) = C 106
This expression doe t explicitlno s y take accoun chemicae th f o t l for i whicmh h mose materiath f I tvalue o restrictiv s transportedAL e i l se us th e f th o e t bu , recommende e ICRth Py b densure s e thamosth t t restrictive situatios i n considered. Wher likele eth y degre f solubiliteo chemicae th r yo l form (organic r inorganico particulara f o ) radionuclid s knowni e e degreth , f pessimiso e m introduce thiy db s procedur quantifiee b n eca referency db tabulatee th o I et dAL values".
s importani t I t that consideratio s givei n potentiao t n l accident scenario i which s e th h chemical for changn mca e through reactions wit surroundine hth g environmen materialsd tan . The most pessimistic value for the ALI should be adopted.
case I nth enrichef eo d reprocessed uranium, such chemical specifiee formb n sca n do the basis of the fuel cycle flow as indicated in Figure 2, and UF6 and UO2 can be cited as materials which are frequently transported. Other possible forms include UO2(NO3)2 UO2F2, UF4 and UO3. According to ICRP 30, among these uranium compounds, those whose inhalation clas(water-solubleD UOs d si an thos 2d 6 F 2an f ,inhalatio UF eo e ar ) n classW (moderately soluble) are UO3 and UF4. To provide an example calculations were applied to a composition arising from a hypothetical cycle:
Initial enrichment of unirradiated uranium 5.0% 235U
Burnup 55GW-d/tU
Uranium reprocesse re-enriched dan d 5.0% 235U s Bq/10 gU x 5 Specific activity
4 5. f (inhalation clas ) sY
isotopie Th c compositio bases nwa d upo burnue nth p calculation code ORIGEe th d Nan enrichment code CENTIS.
When f values were calculated using ALI values of uranium nuclides corresponding to inhalatio nindicate classeW r o resulindicates e a ICRD s n th d, i , is t = P30 Tabl n di f ( eV 0.46- 0.62).
For soluble or moderately soluble compounds of reprocessed uranium, by using ALI values of uranium nuclides corresponding to the inhalation class D or W indicated in ICR , thosP30 e uranium compound classifiee b n ca s LSA-II s da . This is also shown for uranium hexafluoride under the ASTM C-996-90 Standard (enriched reprocessed uranium: ERU) in Table VI.
34 TABL . EXAMPLEV REPROCESSEF EO D URANIUM CLASSIFIE LSA-IS DA I (CHEMICAL FORM OF URANIUM IS CONSIDERED)
Uranium Uranium Inhalation classD Inhalation classW isotopes specific 3 activity ALIb Activitn yi ALIb Activity in [Bq/] U g [Bq] U/ALg 10m I [Bq] 10 mg U/ALI 232JJ 2.24 x 103 4 10 X 8 0.028 1 x 104 0.022 228-rn 2.24 X 104c 4 x 102 0.559 2 10 4x 0.559 234U 3.93 x 10s 5 X 104 0.079 3 x 104 0.131 235JJ 4.00 x 104 3 10 X 5 0.001 " 10 x 3 0.001 a6U 4 10 6.9 x 6 4 10 X 5 0.014 4 10 x 3 0.023 238JJ 5.22 x 104 s 10 x 5 0.002 3 x 104 0.004 TOTAL 5.22 X 105 - 0.68 f2= value , 0.74 valuf 0= e Remarks Uranium compound and f value Uranium compoun f valu d edan
UF6: 0.461 UO3: 0.616 UO2F2: 0.527 UF4: 0.561
a Results calculated by ORIGEN and CENTIS code. Assumptions for calculation: Initial enrichment: 5.0% Reactor type: PWR GW-d/5 5 U t Burnup: Specific power: 38 MW/t U Enrichment of reprocessed uranium: 5.0%. b Valu f ICReo (1977)0 P3 t valubu ,228 f eo T inhalatios hi n clas (minimum)sW . equilibriun I c m with 232U parent.
TABL . EXAMPLEVI URANIUF EO M HEXAFLUORIDE CLASSIFIE LSA-IS DA I (CHEMICAL FORM OF URANIUM IS CONSIDERED)
Uranium isotopes Uranium specific Inhalation class D activity" b [Bq/g U] I AL Activitn yi 10 mg UF6/ALI 232TJ 4 10 4.1 x 4 8 x 103 0.035
238Th 4.14 x 104c 4 x 102 0.700
234JJ 4.62 x 105 5 x 104 0.062
235TJ 4.00 x 103 4 10 x 5 0.001
4d 4 236U 8.40 x 10 5 x 10 0.011
4 4 238U 10 1.1x 5 5 x 10 0.002
TOTAL s 10 6.4 x 4 - 0.811
a Valu enrichedf eo reprocessed uranium, enrichmen AST% t5 M C-996-90 "Standard Specificatio Uraniur nfo m Hexafluoride Enriche 235leso dt % U"s5 . b Value of ICRP 30 (1977), but value of 228Th is inhalation class W (minimum). c In equilibrium with 232U parent. A Estimated value (not specifie ASTy db M Spec).
Next page(s) left blank 35 AppendiI xV MEASUREMEN LEACHINF TO G RATI URANIUF OO M OXIDES
1. PURPOSE
To measur leachine eth g rati f uraniuoo m oxides into demonstrato watert d an , e that these can be are classified as LSA-III.
. EXPERIMENTA2 L
(1) Test samples
UO2 pellet ..... Pellet and fragments (to simulate the broken pellet in the accident condition) UO2 powder U3O8 powder
Test samples were supplied fro e fue) listemth D l n d i manufacturer an C , B , (A s Table numbee VIITh .eac n f i sample o rh o categorytw s si .
TABLE VII. TEST SAMPLES
UO2 powder U3O8 powder UO2 pellet pellet fragment A A — — B B B — — — C C — — D D
(2) Test method and measuring items
(a) Test method
The test method is given in para. 603 of the IAEA Regulations. The detail is shown hi the Attachment I to this Appendix.
(b) Measuring items
(i) Leaching water (demineralized water)
Conductivit t 20°Cya (mS/mH p . d an )
37 (U) Samples
3 U02, U308 powder U(%), O/U , F(ppm) l (ppmC , ) Average grain size (/zm) Specific surface area (m2/g) Specific activity (Bq/g U)
UO2 pellet Specific activity (Bq/g U) (in) Leaching ratio
- Liquor quantity. - Sample weight. - U density in filtrate.
3. TEST RESULTS
tese tTh result showe sar Tabln ni e VIII.
The results show that UO2 powder, U3Og powder and UO2 pellet have lower leaching ratios than the acceptable limit of ERU (enriched reprocessed uranium) of ASTM (Attachment 2).
. CONCLUSIO4 N
tese tTh result demonstrates powder 2 thaUO t , U3O 8pelle2 powdeUO t d meean re th t leaching criterio day7 LSA-IIr f lesn o i sfo s2 thaA 1 I n0. material .
3 only2 UO . r Fo
38 TABLE VIII. LEACHING TEST RESULT POWDER SFO R (UO2, UPELLED 3O8AN ) T
Sample UOg Powder U3 OB Powder t le l UQPe g 2 PelleU0 t frag«enf Item Bt By Ai__Aî_ C, Di Dt C, D, D Height JJLL 10.416 10.116 9.947 10.022 10.083 10.091 10.058 9.868 7.838 7.846 9.197 9.153 10.063 10.093 9.012 8.798 10.221 10.527 FiUrateU 7.54 3.14 19.4 20.2 38.4 40.9 14.9 14.9 2.91 3.66 x10-> x10-J 0.5 0.5 0.106 0.121 0.7 0.6 0.197 0.212 Leached U 1.94 2.02 3.84 4.09 1.49 1.49 2.91 3.66 7.54 3.14 1.06 1.21 1.97 2.12 (K) X10-» XlO-' X10-' X10-' X10-* X10-1 5x10" 5x10" x10-' X10-» 7x10-' 6x10-« X10-» Leached U 2.89 3.71 9.62 4.00 5.44 5.46 1.05 1.20 LU 6.82 1.93 2.01 (%X5amplc) 0.019 0.020 0.039 0.0410.015 0.015x10-' x10-J X10-* X10-' xlO" X10-« X1Q-4 xlO" x10'4 x10-4 x10-4 Conductivity (mS/rn) 0.12 0.59 0.12 0.59 0.59 0.07 0.01 0.07 0.01 PH 6.4 6.6 6.4 6.6 6.6 5.0 5.7 5.0 5.7 Temp. C C) 20 20 20 20 20 20 20 20 20 U (%/Sample) 57.77 87.82 84.67 4.79 0/U 2.05 2.05 (pça pm/u) 32 < 5 ( p pm/u) < 2 <5 < 2 < 5 Mean Diameter 0.75 0.93 2.64 1.98 Surface Area (ni/ g Sample) 3.I 1.93 3.99 0.69 Specific Activity 8.24x10* 7.51X104 5.81X104 7.51x104 7.51X104 8.01X104 B. 16x104 8.01X104 B.IGxlO' BNP L Data (%/Samolc) 0.005 . 0.0014 6.6x10- 4.5x10-' 6x10" 3x10-' Allowable 100 Kg -U0> 100 ko-Uj 0 . -UOg ) K 0 2 jx 45 PH ( R ( 200 kg -U0> x2 Leaching Ratio ASTH ECGU (5X : 0.8) 5 ASTH ECGU (5X ): 0.85 ASTH ECGU (SX) : 0.094 ASTH ECGU (5X : )0.2 1 (lO-'A, ) [RU (5ï : 0.1) 5 ERU (SX : 0.1) 5 ERU (5X) : 0.017 ERU (5X) : 0.039
Refe* Anneo t r I I x Attachmen Appendio t 1 t I xV TEST METHO MEASUREMENF DO LEACHINF TO G RATIO
UO powder SAMPLES 2 U3O8 powder UO2 pellet (pellet and fragment)
Powder: approx. 10 mg, measured precisely MEASUREMENT Pellet: 1 pellet (approx. 10 mg), measured precisely Fragments: , approxmeasuremg 0 1 . d precisely Vessel: glass flas f approxl volumko m 0 20 e. Blank: run in parellel to the main experiment
Demineralised water, 100 ml accurately x Shakmi o et SHAKING I LEFT TO STAND days7 , liquor temperature: C 20°2° C±
FILTRATION Membrane filter (pore size: 0.2 |o,m, nitrocellulose type) 1 SAMPLING 5.0 ml accurately
QUANTITATIVE By radionuclide analysis, absorption spectrophotometry, ANALYSIS fluorometric analysis or ICP spectophotometry
40 Attachmen Appendio t t2 I xV CALCULATIO LEACHINF NO G RATIO CORRESPONDINO GT "0. VALUE2 1A "
. Equatio1 r leachinnfo g ratio
equatioe Th convero nt t "0.1A2" leaching criterion into leaching ratio. (L%/sample followss a s i ) : Wl ) (% — — x 0 10 = L W
where : LeachinWl , g quantity
Wl = O.lAj/Q Q : Specific activity (Bq/g U)
: WeighW f contento t f packagso UOg e( 2) Then, leaching ratio (L%/sample: is ) A L = 10 x ——7— (%) Q x W
. Example2 f calculatioso n
2.1casn I f 5.0. eo % ECGU (ASTM Standard)
9 A2 value: 1.1 X 10 Bq 5 Bq/10 U g X Specifi3 1. c activity:
(1) Package of uranium powder (W = 100 kg UO2): L = 0.85% bdls)2 0.094= X L : 2 % UO g ) k fuePackag R (2 0 l 45 bundlePW f = eo W s( (3) Package of BWR fuel bundles (W = 200 kg UO2 x 2 bdls): L = 0.21% 2.2cas n (ASTI 5.0U f . eo %ER M Standard)
8 A2 value: 8.8 X 10 Bq 5 Bq/10 U g X 7 Specifi5. c activity:
(1) Package of uranium powder (W = 100 kg UO2): L = 0.15%. (2) Package of PWR fuel bundles (W = 450 kg UO2 x 2 bdls): L = 0.017%. (3) Package of BWR fuel bundles (W - 200 kg UO2 x 2 bdls): L = 0.039%.
Next page(s) left blank 41 Appendix VII TREATMEN URANIUF TO M POWDER REGULATIONSE TH N SI
agrees experte th wa endorsed y t dI b san SAGSTRAy db M thaRegulationse th t y ma interpretee b follows da s with regar insolublo dt e material.
(1) Insoluble compounds (UO2, U3O8) may be classified as LSA-III within the existing Regulations by demonstrating compliance with the leaching criteria, which is subject to further investigation.
(2) For dispersible materials (UO2 and U3O8 powder), the application of the containment requirement for a fissile material package to prevent any loss of material under accident conditions should be considered.
A contained material classifie LSA-IIs da stipulates Ii transportee b o dt IP-n a s 2r da o IP-3 type package in accordance with Table V and para. 426 of the Regulations:
materiaA "426 LS SCOd .an l , excep otherwiss a t e specifie paran di . 425, shall be packaged in accordance with the package integrity levels specified in n suci Tabl a , hmanne V e r that, under conditions e likelb o t y encountere routinn di e transport, ther eescapo wiln e contentf lb eo s from packages r wil no ,y losl f shieldinan thero s e b e g affordee th y b d packaging. LSA-II material, LSA-III material and SCO-II shall not be transported unpackaged.
TABLE V (of the Regulations). INDUSTRIAL PACKAGE REQUIREMENTS FOR LSA MATERIAO SC D LAN
Industrial package type Contents Exclusive eus Not under exclusive use LSA-r Solid IP-1 IP-1 Liquid IP-1 IP-2 LSA-II Solid IP-2 IP-2 Liquid and gas IP-2 IP-3 LSA-III IP-2 IP-3 SCO-P IP-1 IP-1 SCO-II IP-2 IP-2 a Unde conditione rth s specifie paran di . 425, LSA-I materia transporte e SCO-d b y an l ma I d unpackaged". Reprocessed uranium dioxide pellete sintereth n i s d solid state (includin a fue n lgi assembly) falling into the category of LSA-III can be transported as a Type IP-2 or IP-3 packag accordancn i e e wit e requirementhth controld an s r transporfo s t stipulate thin di s section. Insoluble reprocessed uranium whose specific activity exceeds lO^Aj/g but does not 3 excee x 10" d2 fore Ath f uranium 2o /gn i , m dioxid r triuraniuo e m octoxide powder,
43 requires further regulation unde rs potentia i LSA-II o t e ldu I dispersibility. Applyine gth conclusion (2) mentioned above, the additional requirement that no material is dispersed from a package under the conditions likely to be encountered during normal conditions of transport and in an accident was added to those for Type IP-2 or IP-3 packages. Non-dispersibility is secured by the additional requirement of demonstrating no dispersal of radioactive material fro mdamagea d packag defines ea paran i d . 564(b)(i) transpore Th . t safet f reprocesseyo d uranium oxide powder can then be deemed as good as or better than that of LSA-III material in a Type IP-2 or IP-3 package.
44 Appendix VIII
A2 OF MIXTURE
3 TABLE IX. EXAMPLE OF CALCULATING A2 VALUE (MIXTURE)
Uranium Content ratio Specific Activity of Fraction of A2(i) isotopes of activitf yo reprocessed activity [Bq] /(O reprocessed isotopes uranium f(i) uranium [Bq/g] [Bq/g U] A2(î)
azU x IQ' 7 82. 8.27 X 10" 4 10 2.2 X 4 0.043 3 x 10' 1.43 X lO'10
234 u 1.7 x 10-3 2.31 x 10" 3.93 x 103 0.752 1 x 109 7.5 x 10-'2 °
235JJ x 10- 0 25. * 10 8.0X 0 3 10 4.0 X 0 0.008 urdimit 0
2 6 4 9 10 236U 2.9 X 10- 10 2.4 X 0 10 6.9 X 6 0.133 10 X 1 1.33 x K)'
Z38JJ x 10- 2 '9. ' 10 1.2 X 4 4 10 1.1 x 4 0.022 uni unit 0
Z28Th - 3.04 X 1013 "4 10 2.2 x 4 0.043 unlimit 1.0 x 10-'8 °
1.0 - 5.00 x 103 1.000 - 1.1 x lO'4 9
8 8.8= A10 20 X
Results calculated by ORIGEN and CENTIS codes. Assumption for calculation: Initial enrichment: 5% Reactor typeR PW : Burnup GW-d/5 5 : U t Specific power: 38 MW/t U Enrichment of reprocessed uranium: 5.0%. b 22!8! Tassumes hi equilibriue b o dt B2mo t U.
45 Appendix IX A VALU URANIUF EO M WHIC SUSPECTEHS I O DT
HAVE2 BEEN CONTAMINATE REPROCESSEY DB D URANIUM
In consideration of the possibility that reprocessed uranium may contaminate unirradiated uranium during processing, ASTM C-996-90, "Standard Specificatior fo n Uranium Hexafluoride Enriche Leso dt s Tha 235% Un5 " defines enriched commercial grade Uranium (ECGU) with the following limits upon uranium isotopes:
uranium-232: 0.002 /xg/g235U uranium-234: 10 000 /xg/g235U uranium-236: 5000 Mg/g235U
concentratioe Thath , is t n limit f thesso e uranium isotopes depenenrichmene th n do t f uranium-235o value calculates Th .f A eo 2wa formul e th y calculatinr db afo Ae 2gth value omixturea f examino t , range eth uraniuf eo m enrichmen classificatior fo t LSA-IIs na .
Table X shows the relation between uranium enrichment and A2 values. ECGU will be transporte fore th uraniuf m o n d i m hexafluoride from enrichment facilitie reconversioo t s n fore facilitieth f uranium o n i d man s dioxide from reconversion facilitie fueo t s l fabrication facilities. Based on the result in Table X, enrichment upper limits for satisfying the LSA-II classification standard (estimated mean specific activit materiaf yo s 10"i l ^ Bq/ r lessgo ) were calculate followss a e b o dt .
Typ f materiaeo l Maximum enrichment
UF6 5.0%
UO2 4.8% As indicate i Appendidh , radioactivxV e impurities were ignored because they make almost no contribution.
However t i mus, e noteb t d that even uraniu f naturamo l origin can i certaih , n circumstances of enrichment, also have a 234U content equal to or greater than 235 9 jig/g0 1000 U (see Table XI) thesr .Fo e natural circumstance valu2 A e 1.1s ei s th 104X , 232 228 236 almost identica e ECGth o t Ul value which includes U, d TUhan . Howevers i 2 A , 'unlimited' for unirradiated uranium to 5% 235U. This calculation demonstrates that ECGU ma consideree yb diffeo dt r insignificantl alsy acceptee ob ma d yan 'unlimites da ' withi2 dA n the specified composition.
47 ê TABLE X. THE RELATION BETWEEN URANIUM ENRICHMENT AND A2 VALUES
ENRICHMENT 4.8% ENRICHMENT 5.0%
Uranium A2(i) ECGU Activity" isotopes [Bq] Spec' [Bq/g] Uranium Rate of f(i) Uranium Ratf eo f(i) [Hg/gU-235] specific activity specific activity
activity f(0 A2(0 activity f(i) At(t) [Bq/gU] [Bq/gU]
232U 3 x 10s 0.002 8.28 x 10" 7.95 x 10' 6.2 5x 1C" 4 2.0 8x lu"' 2 ' 8.210 8x 6.26 x 10"1 2.09 x 10'12 228Th< e " 10 x 4 — 3.04 x 10" 1 10 7.9 x 5 6.25 x lO"4 1.5 610'x 12 8.28 x 10l 6.26 x 10'1 '!.57 x 10'12 234U 8 10 x 1 10000 2.31 x 10" 1.11 x 105 8.7 3x lO' 1 8.73 x lO'10 1.16 x 105 8.7 x IQ'6 1 8.76 x 10'10
235 3 2 3 2 U 00 " 10 8.0 x 0 3.84 x 10 3.0 210'x 0 10 4.0 x 0 ' 3.0 2x 10' 0
236U 1 x 10s 5000 2.40 x 10" 2 10 5.7 x 6 4.53 x 10'3 4.53 x 10'12 6.00 x 102 4.53 x 10'3 4.53 x 10'12
238 4 2 2 U 00 1.210 4x 1.16 x 10" 9.1 210'x 0 1.16 x 10" 8.76 x 10' 0 "Tc " 10 x 9 0.2 6.28 x 10° ° 6.010 3x 4.7 x 410- * 5.2 x IQ'7 17 6.08 x 10° 4.5 910'x 3 5.1 0x IQ' 17 Total — ... 1.27 x 10s 1.00 8.8 1x 10-' ° 1.32 x 105 1.00 8.84 x lO'10
9 9 A2 value A2 vali= ——————————— 6 4 — [Bq10 1.1 ]x 4 1.13 x 10 [Bq] [/-(OMS 2(0
4 Remarks Activity of UF6 8.60 x 10 [Bq/g UF6] Activity of UF6 8.95 x 10" [Bq/gUF6] 5 5 Activit 2 [Bq/1.110 UO 2 x g f yo UO2] Activity of U02 1.17 x 10 [Bq/gUO2]
ASTM C-996-90 "Standard Specification for Uranium-Hexafluoride Enriched to Less than 5% !"U"
IAEA 5557 (As Amended 1990) Table AIH-1. Assumptio radioactivn i s i n thah e"T t equilibrium wit parene hth t nuclid"Ue2 . TABL . VARIATIOEXI F NO WITH ENRICHMENT FACTORS
U5U product ^Uin ^U in enriched product /ig/g^U assay enrichment 23S g/100g U feed pg/g U Percentage U in tails 0.2 0.3 0.4 5 54 8960 9520 10020 58 9620 10230 10760
4 54 8900 9400 9870 58 9550 0 1010 10600
3 54 8770 9230 9630 58 9430 9900 10330
Next page(s) left blank 49 AppendixX
CONTRIBUTIO RADIOACTIVF NO E IMPURITIE REPROCESSEN SI D URANIU VALUE 2 A TH F EO MO T
calculatinn I value gth Af reprocessef eo 2o d uraniu classifo t mm witintt ai yi e e ohon th categoriesA LS e s importanoi th f t i , alwayo t t s take into account daughter nuclidef o s uranium isotopes and nuclides of radioactive impurities (FP, TRU). Whether the nuclides of radioactive impurities (FP, TRU) can be ignored or not is established by taking their representative nuclides int ocalculatioa accordancvalun e i th d Af f eno o 2an e with para4 30 . Regulations.e oth f
The forms and quantities of radioactive impurities (FP and TRU) contained in reprocessed uranium are determined in each case by several factors including the decontamination factor at the tune of its reprocessing. Table XII illustrates that these radioactive impurities make almost no contribution to the value of A2 of reprocessed uranium at levels specified hi national standards.
TABLE XII CONTRIBUTION OF RADIOACTIVE IMPURITIES TO THE VALUE OF A2 (TEST CALCULATION)
ERU (ASTM) ERU (case of Japan) ECGU (ASTM) Nuclideit A: value [Bq]1 Specification' Specific Specification Specific Specification* Specific [Bqf ] gU activity c[Bq/ ] gU activity [Bq/ ] gU activity /lO^Aj /lOX /lO^Aj
Tc " 10 ' x 9 10 x 5 3 3200 2 'x 10- 20 * 27 x 10* 3 6
^g/5 ( ) gU pg/1 0 ) gU 0 (
'«Ru 2 X 10" — — 15 7 5 x W —
H7Np 2 Xn K10 W" 33 ' 10 x 0 5 001 0 001 5 0 x 10*
Pu -a ' 10 X 2 (200 dpm/) gU ' 10 x 0 5 0001
M'Pu 1 x 10'° 0 03 3 0 x 10* — —
2 1 X W 1 4 X 10* 7 10 x 12 TOTAL — (= T value — (=T value for — (="{- valur efo for radioactive radioactive radioactive impurities) impurities) impurities)
IAEA Regulations Safe th er Transporfo f Radioactivo t e Material (1985 Edition) Specification of enriched commercial grade uranium hexafluonde (ECGU) in ASTM C 996 draft and enriched reprocessed uranium hexafluonde (ERU) Enrichment 5% Exampl f discussioeo Japanm n Enrichmen% 39 t
Since irradiated uraniu ms reprocessei refined dan d into reprocessed uraniums i t i , possible tha lattee tth r contains addition i , representativo nt e nuclides show Table th n ei XII, radioactive impuritie TRUd an P ) wit(F s h even smaller contribution. 2 value A th f o eo t s However indicates a , Tablee th n d(b)i d s XIIan , thei) I(a r impacvalue th Af n eo 2t o , specific activity, and the ratio of A2 value to specific activity is so small as to be negligible.
51 TABLE XIII. EXTENT OF CONTRIBUTION OF RADIOACTIVE IMPURITIES AS A WHOLE TO THE VALUE OF A2 (TEST CALCULATION WITH ENRICHED REPROCESSED URANIUM)
(a) All nuclides taken into account Densitf o y Fractiof o n A Effecf o t Nuicl ides act ivib activity 2 )/Xi f( ) (i (Bq/gU) f(i) X(i)(Bq) A2 value a U-232 2. 95 E403 1.33 E-02 3E408 4.43 E-ll 40 E40.9 0 o U-234 14 E40.6 5 7.39 E-01 1E409 7.39 E-10 8. 182E401 ^ U-235 34 E40.1 3 1.41d te E-0i im i 2un 0.00 E400 0.00 E400 •— ' U-236 2.27 E404 1.02 E-01 1E+09 1.02 E-10 13 EiO.1 l •a U-238 9 1E40.1 4 56 E-0.3 2 uni imi ted 0.00 EtOO 0.00 E400 = subtotal 2. 047E405 9. 220E-01 8. 853E-10 9.802E401
52 (b) Only representative nucl ides evaluated Densf o y it Fractiof o n A Effect of idei ic su N act ivy i t b i iv t ac 2 f(i)/X(i) (Bq/gU) f(i) X(lMBq) A 2 value U-232 25 Et0.9 3 1.33 E-02 3E + 08 4.43 E-ll & 4.90 EtOO o U-234 1.64 Et05 7.39 E-01 1EI09 7.39 E-10 8. 182Et01 U-235 3. 14 Et03 1.41 E-02 uni imi ted 0.00 EtOO 0.00 EtOO I U-236 2.27 Et04 1.02 E-01 1E+09 1.02 E-10 13 EtO.1 l 3 U-238 19 Et0.1 4 5.36 E-02 uni imi ted 0.00 EtOO 0.00 EtOO J subtotal 2. 047EI05 9. 220E-01 8. 853E-10 9. 802Et01 &Crt 0 Th-228 1. 54 Et03 6.94 E-03 4E+08 1.74 E-ll 1. 93 EtOO .§ ÏÏ o 3 I— S_O subtotal 1.658Et04 7.465E-02 1.74 E-ll 1.93 EtOO Np-237 2.71 EtOO 1.22 E-05 2E + 08 6. 10 E-14 6. 75 E-03 Fu(a ) 28 E-0.0 1 96 E-0.3 7 2EI08 4.68 E-15 4. 19 E-04 nuc l ide s Pu-241 7.00 EtOO 3.15 E-05 1EUO 35 E-1.1 5 39 E-0.4 4 Tc-99 7. 30 Et02 3.29 E-03 9EI11 3.66 E-15 4. 05 E-04 Ru-106 36 .EtO2 O 1.47 E-05 2EH1 7.35 E-17 84 E-0.1 6 subtotal 7. 431E+02 3. 349E-03 7.257E-14 8. 031E-03 îadioactii e iuiwitie s Total 2. 220E+05 l.OOOE + 00 9. 032E-10 9. 998E-01 A2 = l. 107E9 Next page(s) left blank 53 Appendix XI DAUGHTER NUCLIDE THEI228D F TSO h AN R INGROWTH Accordin e Regulations,th f o parao gt 3 r calculatio30 .fo e valuth f Af o eo n2 a , daughter nuclide with a half-life of either more than 10 days or longer than that of its parent e evaluatenuclidb o t s i ed independently reprocessen I . d uranium ,e uraniu eacth f ho m decan ow isotopey s chaiit s whicn sha i hdaughtea r nuclid generateds ei . Therefore those daughter nuclides must be considered. TABLE XIV. CONSIDERATIO DAUGHTEF NO R NUCLIDES FROM REPROCESSED URANIUM Parent Daughter nuclides Daughter nuclides Feature for Nuclides with shorter with longer calculation of half-lives half-lives A2 values 232 228 U 228Th Th 2M 229^ 225Ra U —— negligible contribution 225Ac of daughter nuclides due to a very small amoun f producto t s 234 u 230Th 226Ra — n 210pb 210p0 235 11 u 23.pa 227Ac — 227Th 223Ra 236 it u 232Th 228Ru 232Th 228Th 238 u 234Th 234U 230Th 226Ra 210pb 2!0p0 Of the many daughter nuclides subject to such evaluation, according to the Table XIV, 228 only Th needtakee b o snt into accoun calculatinn i t value gth Af eo t mus2 I . borne b t n ei mind that 228Th builds up over several years in Table XV [7]. 55 TABL ^T. EhXV INGROWTH3 Years % of equilibrium Activity (Bq/gU) 0.25 0.087 72 0.50 0.166 141 1.0 0.304 248 2.0 0.516 431 4.0 0.766 638 8.0 0.945 787 equilibrium 1.0 828 Activit1 228f yo T h [232U=lQ-9g/g ppb)]1 U( . Therefore e daughtee th effecth f o f i t ,r nuclide b e o 228t Ts i h conservatively evaluated radioactiva , e equilibrium wit s parenit h t nuclide 232U must be assumed. 56 Appendix XII CLASSIFICATION OF AN EMPTY CYLINDER CONTAINING UF HEEL 6 The specific activity of heel in an empty cylinder for uranium hexafluoride is calculated as follows, whic shows hi n wit exampln h6 a befor UF f eeo enrichment. Assumptions Reprocessed uranium PWR 5.0% E, 55 GW-d/t, 38 MW/t, reprocessed 4 years later, evaporated 10 years later day0 3 , s attenuation. cylindeY 48 r Filled amount: 9525 kg UF6 or 6440 kg U: minimum filled amount [8]. Heel r 15.3: maximuo U :6 g 22.63k UF g m8k permissible heel amount [8]. TABLE XVI. CALCULATION OF A, VALUE FOR HEELS Nuclide Activitf yo f(i) A2(i) f(i)/A2(i) A2(Bq) heels (Bq) (Bq/heel) 232U 7 10 4.x 4 0.0011 s 10 x 0 3. 3.7 X IQ'12 8 228Th 1.x 610' ° 0.40 10 X 0 4. 1.6 x 10-' 8 9 234JJ10 x 0 9. 0.022 1.0 x 10 2.2 x 10'" 230 3.4 x 108 0.00084 2.0 x 108 4.2 x 10-'2 9.8 X 10* asTh 1.1 x 107 0.00084 8 10 2. 0x - mu u 2.7 x 108 0.0067 ' 10 1. 0x 6.7 x IG'12 238U3 10 X 9 1. 0.0047 1.0 x 109 - 234Th 2.x 310' ° 0.57 " 10 2. 0x 2.9 x 10-'2 TOTAL 4.0 X 10'° 1.00 - 1.0x 10-2 ' On the basis of the above calculation, the transported item in question is classified as follows: 1. Radioactivity of heel in 48Y cylinder 4-04 x 1Q1° 1.7x 108« UF ) 3 66 22.6 . LSA-I2 I criterion 10"4 AJg = IQ'4 x 9.8 x 108 = 9.8 x 104 Bq/g 57 Since the specific activity under 1 is greater than that under 2 the transported item does not satisfy the standard for LSA-II. Alternatively, to qualify for a Type A transport, the total 8 activit . HoweveryBq mus les10 e tb value 2 ,sx Tabl A tha showI 8 e 9. n ,th totae XV sth l activit muce b o yht greater thax 10. n0 10 Bq thi 4. t sa emptn Ia f y containe transporteds ri contaminatios it , n level mus t exceeno t 3 timed10 s e leveth l specifie r transportefo d d matter exemptee th f do fro1 m42 d Sectionan 7 40 s Regulations (0.04 Bq/cm2 for a radioactive body). If the matter hi question is deemed to be surfaca e contaminated object (SCO) contaminatioe th , n inaccessibls leveit f lo e surface must Bq/cm4 3 Bq/cm10 10 SCO-Ie X r les b SCO-Ie 2o x r lesb 8 o so 2 t o 4 d st I e an ,(botb r hfo the case of a radioactive body). Takin cylinderexample Y gth 48 a internas f eit o , l surface are cm5 1.7s ai 10 d 2 ,0an x its radioactivit 4.0s yi 4x 10 10 Bq/heel. Therefor average eth e contamination leve: is l 4.010x 4 10 2 „, , —————— x 10? 4 2. Bqlcm' = — 1AD 5 1.70 x 10s Henc t doee i f thos t satisfo sno y e an ycategories . In practice the heel is not uniformly distributed following vaporization of the contents but is more concentrated in a band along the base of the cylinder. This is of significance to gamma dose measurements. 58 Annex INVESTIGATION CONTAINER Y TRANSPORE 48 N TH I 6 N SO UF F STO WITH URANIUM FROM REPROCESSED PWR FUEL ELEMENTS Abstrac selected an t d results prepared for BMU, Bonn Alfony b s Tietze* Wuppertal, March 1990 Address . Tietze: A Prof . Dr , . Bergische Universität, Fachbereich Sicherheitstechnick, Gaussstrasse , 56020 0 Wupperta . Telephone1 l : 0202-439-2123. 59 Abstract assessmentthe of determination the The aim is environmentalof dose rates48Y of containers as used for transportation of UF6 taking into account uranium isotope vectors to be expected for reprocessed uranium vectors and the definition of possible countermeasures to reduce dose rates. Results of some measurements of dose rates are given for some examples. Calculation of dose rates are based on the ISOSHLD-programme. The model applied in ISOSHLD-programme was fitted in a suitable manner for calculating dose rates in the environment of 48Y containers. As radiation sources four uranium vectors were used (natural uranium, uranium-composition according ASTM-specification,to reprocessed uranium from PWR-fuel elements with an burnup value of 50 000 MWd (t U)'1, reprocessed uranium from PWR-fuel elements U)'(t with1). burnupa Relevant MWd 000 value27 of daughter nuclides takenare into account container two too. The states "container with homogeneous maximum filling"containerand " homogeneously drained (evacuated)are " considered. resultsThe obtained show: Legal dose rate limit values externalfor radiation exposure are in some cases significantly exceeded. Under certain circumstances drained (evacuated) containers producemay significantly higher dose vicinityratesthe in 48Y ofan container compared with containers filled with natural uranium. If the 48Y container is completely filled with natural uranium, even a certain residual amount of daughter radionuclides from preceding fillings with other uranium vectors lead environmentalmay to dose rates below limit values. Possible countermeasures to reduce dose rates may be: Limitation of time intervals between reprocessing of fuel elements or conversion and transport, intermediate purification of containers according to the special requirements and overpackingsof use the (shieldings) containers48Y of according specialto requirements. 60 . DESIG1 N PROPERTIE CONTAINERY 48 F SO S Basic properties of the 48Y container are compiled in Table 1. A picture of this type of containe gives ri maximuFigure n i Th . e1 containemY 48 conten e th 12.s ri f o t 5t UF 6. TABLE 1. DATES OF THE 48 Y-CONTAINER diameter ...... in.8 (4 )121 m 9m length ...... 3810 mm (150 in) wall thickness ...... 16 mm (5/8 in.) minimum wall thickness ...... 12, 7 mm (1/2 in.) tare weight ...... 235 (520g 9k ) 0lb net weight max...... 12501 kg (27560 lb) gross weight ...... 14860 kg (32760 lb) volume ...... 4, 04 m (142, 7 ft) material ...... Stahl (TTST AST= 9 E2 M A516° 55 oder 60) work pressure ...... 14 bar (200 psig) test pressure ...... 28 bar (400 psig) planned outer pressur epsig2 ...... (2 )r ba 2 5 , 1 . planned temperature ...... -40° 121s C bi ° ,1 ) °F 0 (-40°25 s Fbi accumulation max...... 4,5 Gew.-% U-235) filling limit max ...... 12501 kg (27560 lb) "heer-weight max...... ) lb 0 (5 7 , 22 . cylinder6 UF FIG. L model 48Y. 61 2. MODELLING FOR APPLYING THE ISOSHLD-PROGRAMME programme Th e ISOSHL bees Dha n applie containea o dt r geometry show Figurn i . e2 Dose rates were calculated on the lines " 1 ", "2" ... "7" in certain defined distances from the model surface programme Th . e ISOSHL Dfitte s cylindricai a o dt l geometry. Howevere th , programme ISOSHLD does not allow to calculate dose rates at points in areas as shown in a scraped manner according to Figure 3. The geometrical model therefore had to be modified by "expanding givee "th n model containe "auxiliaryn a y b rA " model containe underB e rth assumption tha tfillee equivalenn botb a y dn hi witma 6 container B h UF t d manneran sA . exampler gainee fo b , dos e 6 y linecalculatine y r poinTh dey b o th ma , rats5 an n o t ea e gth total dose rate, causeradiatioe th y db n diminishe sourcB + dos e eA th ey dratb e causey db source th . DoseB e rate data calculate applyiny b y thin d i programme swa g th e ISOSHLD modifiee th n o d model indicates a , goon i Figurn de i d ar agreemen , e5 t with measurements as demonstrate Figurn di . e5 Dose rate drainer sfo d (evacuated) containers were calculate usiny db gmodea l shown in Figur . Thie6 s mode essentialls i l y assumptioe baseth n do n that daughter radionuclides may be concentrated within a thin layer of 1cm thickness on the inner container surface. This assumptio s use i nr modellin fo d e containeth g o calculatt r e spatiath e l dose ratee Th . comparison of calculated values with experimental data show an acceptably good agreement of data. 381 cm FIG. 2. Geometry of the 48Y container. 62 FIG. Excluded3. areas applicationfor oflSOSHLD. ) (C 258m c 1 381 cm (A) 2200 cm (B) B 122cm FIG. 4. Modified models for applying the ISOSHLD programme on 'excluded' areas. 2t(ia) a pi) »(16) 17(17) 18(18) 21(27) 28(23) 18(19) FIG.Calculated5. measuredand values doseof surface ratesdraineda the at of container48Y (values ^Svin h'1). 63 ) (D m c 1 38 379 cm (E) FIG. Model6. of drained container. . SELECTE3 D RESULTS 3.1. Uranium vectors As sources four uranium vectors were defined. The isotope compositions are figured in Table 2. 3.2. Influence of time on dose rates The concentrations of different daughter radionuclides may be built up, depending on time, resulting in corresponding time depending dose rates at points taken into account. The dose rates in case of vector 1 do not depend on time because mother and daughter radionuclides are practically in radioactive equilibrium. The relative concentrations do not change with time either, Figure 7. In cases of uranium vectors 3 and 4 the dose rates will change with tune as is demonstrated in Figure 8 and Figure 9. 64 TABL USE. E2 D URANIUM VECTORS Uranium vector 1 (natural uranium) U-234 0.005 Mol.- 56 4.916 X IQ'5 g/gU 3 U-235 0.720 Mol.-% 7.10971 X ID' g/gU U-238 99. 275 Mol.-% 9.92841 X 10-1 g/gU Uranium vecto (averag2 r e vector Uranit; Burnup: 27.000 MWd (t.U)'1) 10 U-232 6.5 x 10'8 Mol.-% 6.337 X 10- g/gU 4 U-234 0.0144 Mol.-% 1.416 X 10- g/gU U-235 0.81 Mol.-% 7.99891 X 10-3 g/gU 3 U-236 0.31 Mol.-% 3.0743 X lu' g/gU U-238 98.8656 Mol.-% 9.88785 X 10-' g/gU Uranium vector3 (propose ASTM-specificationn do ) 6 8 U-232 5.1 x lO" Mol.-% 5.000 X 10' g/gU U-234 0.0488 Mol.-% 4.800 X 10-4 g/gU U-235 1.0000 Mol.-% 9.8764 X 10-' g/gU U-236 0.8470 Mol.-% 8.400 X IQ'3 g/gU U-238 || 98.1042 Mol.-% 9.8124 X 10-' g/gU Uranium vecto (thir4 s vecto bees ha r n chosen becaus f calculationeo f Uranitso ; Burnup: 50,000 MWd . (t.U)'1) U-232 x 10' 0 8 3. Mol.-% 2.900 x 10-'°g/gU U-234 0.020 Mol.-% 2.000 x 10-"g/gU U-235 0.700 Mol.-% 6.910 x 10'3g/gU U-236 0.600 Mol.-% 5.950 x 10-3g/gU U-238 98.680 Mol.-% 98.8694 x 10-'g/gU 3.3. Spatial distribution of dose rates Line f constanso t dose rate showe sar Figurn ni throug0 1 e h Figurr containefo 2 e1 r fillings using uraniu respectively3 m d vectoran 2 , 1 s . legae Th l dose rate value somn i e esar cases significantly exceeded. Similar results were gained from model calculations applied to the drained (evacuated) container. Figure 13 gives results for the container previously filled with uranium of vector 1-type. Figurd Figurn I an correspondin5 4 1 e1 e g example givene ar s thesn I . e cases limit value partle sar y significantly exceeded too. 3.4. Influence of uranium composition (uranium vector) The comparison of surface dose rates for all four uranium vectors indicates, correspondin resulto gt s show Figurd Figurn i n an , tha 6 17 e1 et dose rates will grow. Especiall legae th casn yi l4 vector f elimio d an t s3 valu exceedee eb wilt no l d afte year1 r , t wili t l aftebu years4 r comparisoe Th . n between dose rate filler drained sfo dan d containers resulte essentialln di y higher dose rate value drainer sfo d container indicates sa Figurn di e Figurd 18 an , respectivelye19 . 65 1000 f 100 activitAe sth mothef yo d ran daughter nudides remain constant over decades dose ,th e rates will not change within this period year4 s 1 year year6 1 10.2s 3 years FIG. 7. Dose rates for vector 1. 10000 1 year 4 years year6 1 10.2s3 years DoseFIG.8. rates vectorfor 3. 66 10000 1000 100 1 year 4 years 10.23 years 16 years FIG. Dose9. rates vectorfor 4. 1= 6 FIG. 10. Vector 1, container filled, time: I year. 67 O) u c rö 20 nSv/h SO 50 25 liSv/h distanc) (m e FIG. 11. Vector 2, container filled, time: 10.23 years. FIG.Vector12. container3, filled, time: 10.23 years. 68 Öl o c rö +-> i/l "O FIG. 13. Vector l, container drained, time: 1 year. FIG. 14. Vector 2, container drained, time: 10.23 years. 69 FIG.Vector15. container3, drained, time: 10.23 years. 10000 1000 vector 1 vector 2 vectors vector 4 FIG. 16. Comparison of dose rates of the drained container after 1 year. 70 10000 vector! vector 2 vector 3 vector 4 FIG.Comparison17. of dose rates drained ofthe container after years.4 1000 100 vector 1 vector 2 vectors vector 4 FIG.Comparison18. of surface dose filled ratesthe of container after 10.23 years. 71 10000 vector 1 vector 2 vectors vector 4 FIG. 19. Comparison of surface dose rates of the drained container after 10.23 years. 72 REFERENCES [1] INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the Safe Transport of Radioactive Material, Safety Series No. 6, 1985 Edition (As Amended 1990), IAEA, Vienna (1990). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Explanatory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (1985 Edition), Second Edition (As Amended 1990), Safety Series No. 7, IAEA, Vienna (1990). ] INTERNATIONA[3 L ATOMIC ENERGY AGENCY, Advisory Materia IAEe th r Afo l Regulations for the Safe Transport of Radioactive Material (1985 Edition), Third Edition (As Amended 1990), Safety Series No. 37, IAEA, Vienna (1990). ] INTERNATIONA[4 L COMMISSIO RADIOLOGICAN NO L PROTECTION, Publication , Pergamo26 n Press Yor,w OxforkNe (1977)d dan . [5] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Limits for Intake f Radionuclideo s Workersy b s , Publicatio , Partn30 s 1-3, Pergamon Press, Yorw OxforNe k d (1980)dan . ] INTERNATIONAL[6 ATOMIC ENERGY AGENCY, Schedul Requirementf eo e th r sfo Transpor f Specifieo t d Type f Radioactivo s e Material Consignments, Safety Series No. 80, IAEA, Vienna (1986). ] H[7 A YES, T.J., BNFL, Springfields , personaUK , l communication, 1990. [8] US DEPARTMENT OF ENERGY, Uranium Hexafluoride: Handling Procedures and Container Criteria, ORO-651, Rev. 4 (1977). 73 CONTRIBUTORS TO DRAFTING AND REVIEW Consultants Services Meeting, Vienna, November6-8 1989 Corny . F , COGEMA, France Cosack, M. Bundesamt für Strahlenschutz, Germany Goldfinch E. Central Electricity Generating Board, United Kingdom Ha , T.Jyes . British Nuclear Fuels pic, United Kingdom Iwasawa . N , Japan Nuclear Fuel Company Ltd., Japan Kubo, M. Power Reactor and Nuclear Fuel Development Corporation, Japan Selling, H.A. International Atomic Energy Agency Consultants Services Meeting, Vienna, 23-25 March 1992 Dannette, G. COGEMA, France Delplanque, COGEMA, France Hayes, T.J. British Nuclear Fuels pic, United Kingdom Iwasawa N. Japan Nuclear Fuel Company Ltd, Japan Kubo . M , Power Reacto Nuclead an r r Fuel Development Corporation, Japan Nitsche, F. Bundesamt für Strahlenschutz, Germany Ringot, C. Commissariat à l'énergie atomique, IPSN, France Shiomi, S. CRIEPI, Japan Hirata . J , International Atomic Energy Agency Mairs . J , International Atomic Energy Agency Selling, H.A. International Atomic Energy Agency Commentators on the draft TECDOC, Vienna, October 1992 and February 1993 Nitsche . F , Bundesam Strahlenschutzr fü t , Germany Renard . C , Commissaria l'énergià t e atomique, IPSN, France Siwicki, R. National Inspectorate for Radiation and Nuclear Safety, Poland Smith, L. Swiss Federal Office of Energy, Switzerland Svahn, B. Swedish Radiation Protection Institute, Sweden Tsuji . I , Scienc Technologd ean y Agency, Japan Ad-hoc Working Group III of the second Technical Committee Meeting for the Revision of the IAEA Regulations for the Safe Transport of Radioactive Material, Vienna, 17-21 May 1993 Cheshire, R. British Nuclear Fuels, pic, United Kingdom Iwasawa . N , Japan Nuclear Fuel Company Ltd, Japan Ringot, C. Commissariat à l'énergie atomique, IPSN, France Pollog, T. International Atomic Energy Agency 75