Research Reactor Core Conversion Guidebook

IAEA-TECDOC-643

Research reactor core conversion guidebook

Volume 5: Operations (Appendices L-N)

INTERNATIONAL ATOMIC ENERGY AGENCY /A> RESEARCH REACTOR CORE CONVERSION GUIDEBOOK VOLUME 5: OPERATIONS (APPENDICES L-N) IAEA, VIENNA, 1992 IAEA-TECDOC-643 ISSN 1011-4289

Printed by the IAEA in Austria April 1992 FOREWORD

In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety d licensinan g issue conversior fo s f theino r reactoU HE r f coreo e sus froe th m fuel to the use of low enriched uranium (LEU) fuel. Two previous guidebooks on research reactor core conversion have been publishe e IAEAe th firsTh .y tb d guidebook (IAEA-TECDOC-233) addressed feasibility studies and fuel development potential for light-water-moderated research reactors and the second guidebook (IAEA-TECDOC-324) addressed these topic r heavy-water-moderatefo s d research reactors. This guidebook n fivi , e volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor greatle b o ,t y facilitated. This guidebook has been prepared at a number of Technical Committee Meetings and Consultants Meetings and coordinated by the Physics Section of the International Atomic Energy Agency, with contributions volunteerey b d different organizations. The IAEA is grateful for these contributions and thank e expertth s s froe variouth m s organization r preparinfo s e detaileth g d investigations and for evaluating and summarizing the results. EDITORIAL NOTE

preparingIn this material press,the International the for staffof Atomic Energy Agency have mounted and paginated the original manuscripts and given some attention to presentation. The views expressed necessarilynot do reflect those governments ofthe Member ofthe Statesor organizations under whose auspices the manuscripts were produced. The use in this book of particular designations of countries or territories does not imply any judgement publisher,the legalby the IAEA, to the status as of such countries territories,or of their authorities institutions delimitationand the of or theirof boundaries. The mention specificof companies theirof or products brandor names does implynot any endorsement or recommendation on the part of the IAEA. This text was compiled before the unification of Germany in October 1990. Therefore the names German Democratic Republic and Federal Republic of Germany have been retained. PLEASE BE AWARE THAT MISSINE TH AL F LO G PAGE THIN SI S DOCUMENT WERE ORIGINALLY BLANK PREFACE

Volum consist5 e f detaileo s d Appendices L-N, which contai nvarieta f o y useful informatio operatioe th n o n f researco n h reactors with reduced enrichment fuels. Summaries of these appendices can be found in Chapters 12-14 of Volum (SUMMARY1 e f thi)o s guidebook. Appendi containL x sa summar f o y necessary and recommended experiments for reactor startup. Appendix M provides informatio e procedureth n no d experiencean s f severao s l reactor operators with both mixe d fulan dl cores with reduced enrichment fuels. AppendiN x contains informatio n transportationo f boto n h fres d spenan h t fuel elements, n speno t DepartmenS fueU le th storage n f Energy'o o t d ,an s receip d finanan t - cial settlement provisions for nuclear research reactor fuels. e topicTh s whic e addressear h Volumn e appendicei dth , 5 e whicn i s h detailed informatio e summare foundth b n d ca y,nan chapter Volumn i se ar 1 e listed below.

VOLUME 1 VOLUME 5 SUMMARY Topic __APPENDI_ X Chapter

L Startup Experiments 12

Experience with Mixed and Full Core Operation M 13

Transportation, Spent Fuel Storage, and Reprocessing N 14 CONTRIBUTING ORGANIZATIONS

ArgonnL AN e National Laboratory United States of America

Chalk River Nuclear Laboratories CRNL Canada

Commissariat a I'Energie Atomique CEA France

GA Technologies Inc. 6A United State f Americso a

GKSS-Forschungszentrum Geesthacht GmbH GKSS Federal RepublI f Germano c y

Japan Atomic Energy Research Institute JAERI Japan

Kyoto University Research Reactor Institute KURRI Japan

Netherlands Energy Research Foundation ECN Netherlands

k RidgOa e National Laboratory ORNL United State f Americso a

Österreichisches Forschungszentrum Selbersdorf ÖFZS Austria

N T Transnuklear GmbH Federal Repub f Germano c lI y

United Kingdom Atomic Energy Authority UKAEA United Kingdom

University of Michigan - Ford Nuclear Reactor FNR United States of America

U.S. Departmen f Energto y USDOE United States of America

The IAEA Is grateful for the contributions volunteered by these organizations and thanks their expert r preparinfo s e detaileth g d Investigation r evaluatinfo d an sd summarizinan g e th g results presente thin I d s Guidebook. CONTENTS

APPENDIX L. STARTUP EXPERIMENTS L-l. GKSS: Startup experiments with reduced enrichment fuels...... 11 W. Krull

APPENDI . EXPERIENCXM E WITH MIXE FULD DAN L CORE OPERATION Mixed cores M-l. CRNL: CRNL experience with research reactor fuel conversion and mixed core operation ...... 9 1 . R.D. Graham M-2.ÖFZS: Experience with mixed cores in the ASTRA reactor ...... 25 CostaJ. M-3.KURRI: HEU-MEU mixed core experiments in the KUCA ...... 39 Kanda,K. Shiroya,S. Hayashi,M. Kobayashi,K. ShibataT. M-4.ORNL/ANL transitioe Th : n whole-core phasth f eo e demonstratiot na Ridgk theOa e Research Reactor ...... 1 5 . Hobbs, W. Bretscher,R. MM. R.J. Cornelia, J.L. Snelgrove Full cores M-5.FNR: Operational impactenrichmenw lo f so t uranium fuel conversion no the Ford Nuclear Reactor ...... 67 R.R. Burn M-6.JAERI: Ful lfue U corl edemonstratioME JMTe th n Rni ...... 3 7 . M. Saito, Nagaoka,Y. Shimakawa,S. Nakayama,F. Oyamada,R. OkamotoY.

APPENDI . TRANSPORTATIONXN , SPENT FUEL STORAGE REPROCESSIND AN , G N-l. Transportatio f fresno h fuel elements N-l.l. TN: Transportation of MTR fuel elements within the Federal Republic of Germany ...... 85 N-l.2. KURRI: Transportation of fresh fuel elements for Japanese research reactors ...... 93 Kanda,K. NakagomeY. N-1.3. UKAEA: The UKAEA unirradiated fuel transport containers ...... 95 R. Panter N-1.4. GA: Transportation of unirradiated TRIGA-LEU fuel ...... 97 N-2. Transportatio f spenno t fuel elements N-2.1. GKSS : Remark transportatioe th n o s f spenno t fuel elements ...... 9 9 . W. Krull N-2.2. TN: Transnuklear spent fuel shipping containers ...... 103 N-2.3. UKAEA: UKAEA's 'UNIFETCH' irradiated fuel transport containers ...... 7 10 . R. Panter N-2.4. CEA: The transport of spent fuel elements of research reactors ...... 113 N-2.5. GA: Transportation of irradiated TRIGA-LEU fuel ...... 117 N-3. Spent fuel storage N-3.1. ANL: Nuclear criticalit fueU l HE elemen yd assessmenan U t storagLE f o t e ...9 11 . R.B. Pond, J.E. Mates N-3.2. KURRI: Fresh fuel storage ...... 7 13 . Y. Nakagome, K. Kanda, T. Shibata N-3.3. ECN: Spent fuel storage ...... 143 E.G. Mérelle,A. Tas N-4. USDOE: Receip financiad an t l settlement provision nuclear fo s r research reactor fuels ...... 149 Appendix L STARTUP EXPERIMENTS Appendix L-1

STARTUP EXPERIMENTS WITH REDUCED ENRICHMENT FUELS

. KRULW L GKSS — Forschungszentrum Geesthacht GmbH, Geesthacht, Federal Republi f Germanco y

Abstract A summar f necessaro y d recommendean y d experiment r startufo s f o p a reactor with reduced enrichment fuel s providedi s .

1. General remarks

Recommended startup procedures and experiments depend on a number of factors suc: has

completenese th - nucleaf o s thermodynamicad ran l calculations, completenese th - dynamif so d safetcan y related calculations,

comparisoe th actuae - (LEUw th d (HEU ne f lol o n d ))an core design,

- the operation with mixed (HEU + LEU) or only new (LEU) reactor cores, content5 U- w ne , d an d ol -

- changed or unchanged fuel element design,

- change unchanger o d d contro designd lro , - knowledge of thermal flux distribution (power distribution),

- knowledge of burnup values,

trie th p value- safetd san y margins.

Having most of these calculations and the needed knowledge one has to perform startup experiment o reasontw r sfo s

checo t calculationse kth - ,

learo t n- from experiment flus powed sa xan r distribution- re d san activity values will change and as normally the core configuratio r researcnfo h reactor t fixedno s i s.

11 followinge th n I , proposal necessarr fo s d recommendeyan d startup experiment givee sar n together wit hshora t commentary (see also

2. Necessary startup experiments

2.1 Critical experiments for the standard and for modified core configurations

reactorw fe a increase Onl; r th sC. fo y U-23n ei 5 content will compensatinr befo smala onl) % f o y0 l 2 e amounth g 5 (1 t reactivity effects. Normall manr fo yy reasons (e.g. economics) 5 contenthU- e t wil muce lb h higher. Therefore careful critical experiment e necessarsar learo t ye reactivit th n y behaviouf o r the new fuel elements and control rod worths for different core configuration.

2 Neutro2. n flux measurements

) A Local thermal neutron flux distributions

fuen i l. a elements paralle fuee th lo t lplates ,

b. fuel elements plus control fuel elements vertical to the fuel plates,

nea. c r irradiation positions.

C.; The local thermal neutron flux influencing the formfactors will change especially when using mixed cores. One has to be very carefu assuro t l e tha unalloweo n t d formfactory sma occur.

) B Global thermal neutron flux distributions safete th r y Fo analysi ; necessars C. i t si havo t yaddition ei n e locatth o l power distributio globae th n l power distribution wit changes it h differenr fo s t core configurations, controd lro height d burnupsan .

12 C) Neutron spectra for different positions

C.; The enrichment reduction will lead to a harder neutron spectru mann i yd caseman changeo t s corn si e configuration. Thw spectrene d ratioaan necessar e sar planninr yfo disd gan - cussing the experiments in and outside the reactor core.

D) Neutron fluxes and Gamma-fluxes in the irradiation positions

C.; The enrichment reduction will lead to a harder neutron spectrum and in many cases to changes in core configuration. w spectrne ratiod e aan Th necessar e sar planninr yfo d disgan - cussin experimente gth d outsidan n reactoe i s th e r core.

3 Reactivit2. y values

A) Control rod worths, reactivity speed

C.; For different and new core configurations surprising results can be obtained. Normally, limitations on the reactivity speed exis safetn i t y reports when considering startup acci- dents.

) B Reactivity wort somf ho e element reflectod san r elements

C.; These reactivity measurements are of interest for the reactor operato havo t r epracticaa l knowledge when changing core configurations using different type of loops for different position r e.gfo .d replacinsan g reflecto e partth f o s y fueb r l elements.

) C Reactivity wort loopsf ho , rigs, capsules etc.

; ThesC. e reactivity measurement interesf reace o th e -r sar fo t tor operator to have a practical knowledge when changing core configurations using different type of loops for different positions and for e.g. replacing parts of the reflector by fuel elements.

13 ) D Reactivity values when replacing oval type control rods with forked type control rods. contrôlable th s A ; C. e reactivit e controy changon ma y d b yelro by more than 50 %, many new experiences will be obtained. And necessars ii t mako t y e them stepwise.

) E Contro drod lro p time

; Changepossible C. b y s ma othef ei r type controf o s l rode ar s used.

2.4 Fuel elements with thermocouples

Fuel elements instrumented with thermocouples will be very helpful in the licensing procedure. The fuel plate temperature measurements are an additional check of the nuclear and thermo- dynamical calculations. Such instrumented fuel elements are only necessary during startup and first fuel cycles.

) A Actual fuel plate temperatur selecten i e d positions

sure b eo T tha ; tC. nuclea thermodynamicad ran l design calculations are correct, fuel plate temperature measurements for selected position recommendede sar thess .A e wil onle lb y point measurements, specia spot lho t factors use e b hav do t ewhe n comparin measuree th g d temperatures wit theoreticalle th h y allowed temperatures.

B) Loss of flow experiments with different trip values and for different experimental conditions (position of the instrumented fuel element, different failure considerations)

C.; Of interest is the temperature behaviour between stopping the primary coolant and the scram of the reactor and the flow inversion. Of importance for the first case are the trip values and for the second case the considerations from the failure tree especiall r thosfo y e reacto w neverno o roperatort p u o swh performed such experiments.

14 2.5 Coolant flow distribution

C.; Of main importance if the fuel element geometry and the control fuel element geometry will be changed.

Recommende. 3 d startup experiments

3.1 Isothermal temperature coefficient

C.: Necessary for safety calculations. Measurements, e.g. by cooling the primary water.

2 Powe3. r coefficien reactivitf o t y

C.; Of interest for the reactor operator and easy to measure.

3.3 Pressure drop

C.; Of main importance if the fuel element geometry and the control fuel element geometry wil changede lb .

3.4 Criticality of fresh and spent fuel storage ( 0,95)

C.; If a higher U-5 content will be used one has to assure that k 0,95 in any case. Measurements e.g. with pulsed neutron technique.

3.5 Reactivity values for Xe-equilibrium, Xe-peaking and burnup r lonFo g : terC. m fuel cycle planninrestartinr fo d gan ga reactor after a scram, these reactivity values for actual core configuration greaf o e t ar simportanc e reactoth o t er operator.

Finall. 4 y

More detailed recommendation needee th r d fo sstartu p experimentn ca s only be given for an actual case.

15 These recommendation e influencesar experiencee th - y re b d a f so search reactor operator. From other standpoints (designer, theorist, independent experts or safety authorities) more or less of other startup experiment e recommendeb y sma r requiredo d .

REFERENCE

III IAEA-TECDOC-304 Core Instrumentation and pre-operational procedures for core conversio LEUo t U , IAEnHE A 1984

16 AppendixM EXPERIENCE WITH MIXE FULD DAN L CORE OPERATION

Abstract

Experience with mixe d fulan d l U fuecoreHE lf o s and reduced enrichment fuel e describear s r fo d several reactor designs that range from a coupled- core critical facilit o higt y h power reactors with both rodde d plate-typan d e fuels. MIXED CORES

Appendil xM-

CRNL EXPERIENCE WITH RESEARCH REACTOR FUEL CONVERSION AND MIXED CORE OPERATION

R.D. GRAHAM Chalk River Nuclear Laboratories, Atomic Energy of Canada Limited, Chalk River, Ontario, Canada

Abstract The NRX and NRU reactors at CRNL, both originally fuelled with natural uranium metal, have undergone a number of fuel conversions before arriving at the present highly enriched uranium-aluminum alloy fuel designs. The history of the changes and the regulatory approvals required to make the changes are summarized.

1. INTRODUCTION

The Chalk River Nuclear Laboratories (CRNL) in Canada have had a considerable amount of experience in fuel conversion and mixed core operation of higo thtw eh power, heavy water research reactors located there. Both NRX, built in 1947, and NRU, built in 1957, were originally fuelled with natural uranium metal. Both reactors have gone through several fuel conversions before arriving at the present design of highly enriched (HEU) uranium- aluminum alloy pin-type fuel loadings.

l majoAl r fuelling changes required some for f safeto m d hazardan y s assessment d approva an ,e appropriat th y b l e regulatory body. e th Non f o e changes create y majoan d r problem r significantlo s y affecte e e safetth th d f o y reactors.

The various fuel changes which occurred and resulting mixed cores, and the regulatory approvals that were necessar summarizee ar y d below, firsr fo t NRX, then NRU.

2. NRX EXPERIENCE e largar eU volumNR d an e Bot X tank-typNR h e reactors D£s i 0 X NR . moderated O cooled,Ho graphitd ,an e reflected. Wite currenth h loadinU HE t g it operates at about 25 MW although it has operated at up to 42 MW in the past. Vertical through tubes in the reactor vessel (tank) form 199 lattice sites in a hexagonal arrangement. Fuel assemblies, shut-off rods, and various experiment e installear s thesn i d e through-tubes, suspended inside individual flow tubes.

19 With the initial U-metal loading, almost all lattice sites other than shut-off rods, were occupie y fueb d l assemblies, althoug mid-1950'e th y b h a s number of Pu-Al or enriched U-A1 alloy assemblies had been installed as "boosters" to free some lattice sites for experiments. By 1958 the loading consiste 0 U-meta14 o t f abou o dl0 fue13 t l assemblies, abou 0 "boosters"3 t , d abouan miscellaneou0 2 t s experiment d isotopan s e irradiations.

Table 1 summarizes year-end reactor loadings from 1958 through 1981. Plans to convert NRX to natural IK^ fuel were begun in the late 1950's and in 196a tes0 t irradiatio 0 UO5 ^ o t fuef som o n0 l4 e assemblies, installen i d place of U-metal fuel assemblies, was begun. No formal safety and hazards assessment was performed at this time, but the change was reviewed extensively in the normal process for approval of experimental irradiations.

Beginning in 1962 a gradual conversion to a reactor loading with natural UO, fuel plus 7 pin U-A1 alloy (93% U-235, 0.90 g/cm3) was carried out in order to increase the neutron flux in the reactor and provide more room for experiments. A safety and hazards assessment of the new fuel loading was carried out and approval for the change obtained from the regulatory body. e loadinTh g after this change consiste abouf o d 0 U(>8 t 2 assemblies 7 0 U-A( ,5 1 pin) assemblies, 30 miscellaneous experiments and isotope irradiations, and a number of vacant positions.

Between 1968 and 1970 the UU£ fuel assemblies were gradually removed leavin reactoe th g r fuelled entirely wit % enricheh93 d U-A1 fuel abou0 7 t assemblies sane th e t timA e therma. th e le reacto poweth s f reduceo rrwa d from 42 MW to 30 MW while maintaining the same neutron flux levels. Coincident with this chang completa e e safet d hazardan y s assessmene th f o t NRX reactor was performed and approved by the regulatory body.

A further change occurred between d 197197an 4 2 whe slightla n y different design of 7 pin U-A1 alloy fuel assembly was introduced. The new design featured a thinner cladding, smaller flow area, and a longer fuel length (2.74 m versus 2.44 m). This required an addendum to the previous safety and hazards analysi d regulatoran s y approval. Sinc ereactoe 197th 4s remaine ha r d entirely fuelle n U-A 2.7y m lonpi 1b d4 7 gfue l assemblies (93% enriched), althoug e numbeth h assemblief o r s variedha s .

Note tha te abov th eac f eo h change s madswa e graduall o thae s y th t reactor in fact went through a series of mixed loadings intermediate between the old and new. At no time did any of the loading changes cause any significant problems in the operation of the reactor.

3. NRU EXPERIENCE

reactoU ThNR e ^0s i r moderate ^n a 0 s cooled,d reflectorha an d d an . Whil t initialli e , witye currenMW th operateh 0 22 t t a loadind t operatei g t a s 125 MW. Unlike NRX, the NRU reactor vessel (tank) does not have through tubes, but has a hexagonal array of 227 lattice sites formed by tubes extendin ge tan th upwardk f o throug p sto e uppe froth he th rm shielde th f o s reactor. Like NRX, the assemblies installed in each lattice position are suspended within their own individual flow tubes.

Initially NRU was fuelled with about 190 natural uranium metal fuel assemblies each consisting of five 3.05 m long plates (or "flats") in an aluminum flow tube. During the early 1960's some HEU assemblies were installed, primarily as test irradiations. Table 2 summarizes NRU year-end loadings from 1981o 196t 1.

20 TABLE 1: NRX mixed core history

Cor f eCalandeo Loadind En t ra g Year Number of Assemblies of Each Type Occupying Lattice Sites

00 rH rH rH 3 0) <0 0) c •0 10 i-i 3 rH o 8-8 (K CO W f*-* ^ fu 8 3jj §1 rH ^^ 00 r-j "tf rH *J u 1 g 3 to Year •H tu rH i **£ • 1 3 •H * •O tJ 0 * rH rH G X C/) 13 CM S CN s o tu 13 cd O V— ' -5 [v] 4J 4J Q) XJ t/3 rH rH C **^- ^J • Ï- CO G * 0) CM CU i-jr-j < rH •H D •S» (0 4J *" • p as o S CM 0 3 c < Q* m i P-. H Qj U fl) fi, O T~J CJ O 1 1 3 3 3 . -a US rH D S S tu Cu U, pL£ rlS 3 MC K> <

1958 136 1 _ 6 22 5 19 10 0 1959 128 4 1 17 13 4 22 10 0 1960 86 43 - 29 - 5 26 10 0 1961 91 42 - 29 - - 6 21 10 0 1962 26 41 31 16 - 24 - 7 21 10 23 1963 15 38 34 - - 52 - 8 26 10 16 1964 8 32 47 52 - 5 31 10 14 1965 1 24 59 51 3 9 28 10 14 1966 - 23 59 53 - 8 28 10 18 1967 - 7 63 40 13 9 27 10 30 1968 - - 44 45 14 8 23 10 55 1969 - - 20 60 5 9 29 10 66 1970 - - - 66 3 5 21 10 94 1971 65 4 9 23 10 88 1972 36 36 9 31 10 77 1973 9 63 11 31 10 75 1974 - 78 10 27 10 74 1975 77 8 24 10 80 1976 81 7 22 10 79 1977 54 6 15 10 114 1978 56 5 19 10 108 1979 52 3 18 10 116 1980 50 2 13 10 124 1981 60 1 14 10 114 Fuel Assembly Designs - NRX

U-Metal Fuel: Natural U cylinder - 34.5 mm ÖD x 3.06 m long AI clad (2 mm)

U02 Solid: Natural UOo pellets - 35.8 ram ÖD x 3.05 m long (total) I claA d - (1.27 mm)

UOo Annular :x 3.0 D 15.x I 5 m D m 3Ö m Natura m m o pellet8 . UO l5 3 - s long (total claI A d)- (1.2) 7mm

U-A1 Slug (HEU); Ü-A1 allo 93y- 20.2 - U-23U 0n 5i g/cm 3 U-23claI A d5- (2 mm) - 12 slugs 34.5 mm ÖD x 203.2 mm long - 25.4 mm AI spacer between each

Pu-AI Slug: Pu-Al allo 0.1y- slug2 01 g/cm- s u P 3 ) mm l clalon m A 2 ( ra d- g 0 23 x D Ö 34. m m 5

7 Pin U-A1: U-A1 alloy - 93% U-235 in U - 0.90 g/cm3 U-235 - cluster of 7 pins - (a) 6.35 mm OD x 2.44 m long cladl - 1.1A m 4m . 2.7x D ) O 6.34lonm m (b 5m g - 0.76 mm Al clad.

Fast Neutron: Annular fuel rod with dry central cavity - may be or U-A1

Vacant or blocked lattice sites are generally peripheral, low flux positions, althoug w higfe h a h flux position e blockeo tubt ar s ee leakagedu d .

21 TABLE 2: NRU mixed core history

Core Loading at End of Calander Year Number of Assemblies of Each Type Occupying Lattice Sites

CO CO CD CO Pu I-H co PU iH CO fK i-H CO P^ t—f CO P-( iO —C 1 u >; 4-1 4-i cn m u s 3 frj C H 01 CM (U CM 01 CM a) CM a) CM m TB cd Q.T3 M TC 3 T3 u o •H sa CM 3 1 PO 1961 192 4 - - 3 18 10 1962 187 5 - 5 - 1 4 18 7 1963 184 6 5 2 - 1 4 18 7 1964 - - 33 67 - 10 17 22 78 1965 - - 9 - 75 - 11 30 22 80 1966 - - - 81 - 12 49 22 63 1967 - 78 4 - 12 54 22 57 1968 - 81 - 13 50 22 61 1969 - 58 29 15 42 22 61 1970 - 86 14 68 22 37 1971 - 83 10 67 22 45 1972 1973 Shutdown for Vessel Replacement 1974 - 73 6 49 22 77 1965 16 64 10 46 22 69 1976 54 28 12 44 22 67 1977 7 79 10 43 22 66 1978 1 86 9 44 22 65 1979 - 88 10 41 22 66 1980 88 9 42 22 66 1981 88 10 42 22 65 Fuel Assembly Designs - NRU

U-Metal Fuel: natura5 metalU l platest a 3.0 2 5lonm ( - g t a 1 ; mm 3 4. 49.t a x m 2 8m ; mm 5 4. x m tn 31.1 54.5 mm x 4.3 mm) - Al clad 0.62 mm

Double Annular HEU: U-A1 Alloy - 93Z ü-235 in U - 0.36 g/cm3 U-235 - 2 concentric tubes 3.05 m long - inner x 1.7 thick n D O 35.8ra m ,6m oute x 1.5 r D m 2O 52.m m 3m thick - Al clad 0.76 mm

Single Annular HEU: U-A1 alloy - 93Z U-235 in U - less than or equal to 0.36 g/cm3 U-235 - Fuel annulu 0.7x s thicm D 6n 51.O 2.7m kx 5n 4lonm g

12 Pin U-Al: U-A1 U-232 alloU 93 n 5i y- pin2.72 x 1 D 5.4s- 4lonm O - l A clam - g 8m n i d (0.7) mm 6 - various U-235 densities have been use varyiny b d g alloy content: 0.18 g/cm3 0.27 g/cm3 0.39 g/cm3 0.50 g/cm3 0.63 g/cm3

Fast Neutron: 15 fuel pins form an annulus with a dry central cavity - fuel may be U02 or Th02~U02

Flux Peaking: Used to increase neutron flux in certain regions of core Th0n pi 2 -U09 1 2- Vacan blocker o t d lattice site e generallar s y peripheral w flu,lo x positionf o s limited value.

22 During 1964 the reactor was shutdown and converted directly to highly enriched (HEU) U-A1 alloy fuel e reactoTh . r thermal powe s reducerwa d from whilW M 0 e7 maintainino t W M 0 same 22 th ge neutron flux levels A complet. e safet d hazardan y s assessmen reactoe th s carrief t thio twa a r st ou dtim d an e regulatory approval was obtained. Initially two types of U-A1 fuel assembly (about 100 assemblies in total) were installed; an annular U-A1 alloy design

with Ü-235 densit f abouo y t U-An 0.3pi 162 allo1 g/cm a yd 3,an desig n with 0.18 g/cm 3 U-235. As experience was gained with operation of the enriched

reactor, the fuel was gradually changed to the 12 pin design (about 80 assemblies) with a slightly higher U-235 density (0.27 g/cm)3 . This change had been addressed and approved with the initial safety and hazards assessment.

Between 196 d 197 6e reactoan 0th r loadin s changegwa d three more times. Each change simply involved an increase in the U-235 content of the U-Al fuel by increasin quantite th g f uraniuo alloye y th e U-23 n Th .i m 5 density increased to 0.39 g/cm3 in 1966, to 0.50 g/cm3 in 1968 and to 0.63 g/cm3 in 1970 r eac Fo thesf .o h e change n addendu a spreviousle th o t m y approved safety and hazards assessmen e reactoth requires f o trwa d regulatoran d y approvas lwa obtained fuee th l s loadinA . g increased, reactor powe gradualls rwa y . increaseMW 5 12 o t d Note that these later changes occurred gradually with the reactor loading going throug numbeha f intermediato r e mixed stages .e loadin th Non f o e g changes cause significany an d t problems.

23 Appendi2 xM-

EXPERIENCE WITH MIXED CORES IN THE ASTRA REACTOR

. CASTJ A Österreichisches Forschungszentrum Seibersdorf GmbH, Seibersdorf, Austria

Abstract Core configurations compose f differeno d t MTR-type fuel elements were operated in the ASTRA-reactor. A thermal hydraulics analysis s performewa y calculatinb d e safetth g y margins. Resultf o s measurement f LEU-teso s t element e ASTRA-corth n i s e presentedar e .

1) Introduction

The ASTRA-reactor, a pool type research reactor with a thermal power of 8 MW, is in operation since 1960. In the course of reactor operation several modification MTR-typf o s e fuel elements have been used. Reasons for the change of the fuel element design were mainl e incentivth y o improvt e e performancth e e corth e f eo decreaso t d cose an th f ereactoto r operation. Late n concerno r s on proliferation became important e mai.Th n changee fueth l f o s element design affected element geometry numbee ,th f platero s per fuel element, the amount of uranium per plate and the enrichment of fuel. New fuel elements with different design were loaded into the existing core gradually step by step so that mixed cores arose and wer operation i e longea r fo nr time period o importan.Tw t mixed core configuration describee ar s n thii d s paper e firs.Th t type of mixed core configurations described is related to the transition phase from curve o straight d t fuel element geometry, the second type is related to the test of LEU-fuel elements in the ASTRA-core mixel al dr cor.Fo e configuration thermala s - hydraulic safety analysis is performed.

2) Fuel Element Types

Table 1 gives a description of the essential characteristics of all fuel element ASTRA-reactoe s th use n i d r since starf to reactor operatio e seeb n 1960n i n ca froe tabl.s th A me th e most important change ocurred in 1969 when the geometry of

25 TABL . FUEE1 L ELEMENT DESCRIPTION

Yea f Delivero r y 1960 1965 1969 1974 1982 1982

Plates/Standard F.E. 16 19 23 23 23 20

Plates/Control F.E. 7 9 17 17 - -

Shap f Plato e e curved curved straight straight straight straight

Outer Plates Aluminium Fuel Fuel Fuel Fuel Fuel

Plate Thickness (nun) 1.27 1.27 1.27 1.27 1.27 1.60

Water Channel Thickness (mm) 3.12 2.95 2.23 2.23 2.23 2.44 (2.12*) Uranium Enrichment (%) 90 90 90 90 45 20 (93*) Fuel Meat Material UA1 UA1 UA1 UA1 UA1X-A1 U308-A1

U-235 (gJ/Standard F.E. 193 197 263 285 320 350

U-235 (g)/Control F.E. 04 93 184 206 - -

Gar Controfo p d (mmRo l ) 1 x 28.5 x 28. 1 5 2 6. x 2 2 x 6.2 - -

* Data for Control Fuel Element

fuel plates changed from curve o straight d t plates e numbeth , r f plateo r standarpe s d fuel elemen U-23, 23 t 5o t froweigh 9 1 m t e wate th p decrease ga rd an g 3 d 26 fro o t m g 2.9 froo 7 t 19 m5m m 2.23 mm. The change of the control fuel element and the control rod was even yet more important. The central bar-type absorber- rod consisting of boron carbide was replaced by fork-type absorber blades Ag-In-Ca mad f eo d alloy y eliminatin.B e th g central water gap in the control fuel element a strong reduction in flux and power peaking could be achieved.

n 198I e firs2th t test element enriche% wit 0 2 h d uranium, 35fue0 2 U-23g 0 ld platean 5 s loadeswa d inte coreth oe .Th design of this LEU-test fuel element was based on calculations performe cooperation i d n witArgonne th h e National Laboratory e RERTith n R prograe I.A.E.Oth f o m .

3) General Characterization of Mixed Cores

characterizee b A corn ca e mixeds a d f fresi , h fuel elementf so different desig e simultaneousl e loadear ar nd an d y together e coreith n . MTR-type fuel element diffey sma mann i r y aspects as numbe platesf ro , thicknes d shapsan f plateeo , uranium weight

26 per plate, enrichment, fuel meat material, area of meat and so on. But even if fresh fuel elements with the same design and uranium weight are introduced in the core, fuel elements became quickly different if burnup is becoming effective. In the equilibrium core of the ASTRA-reactor fuel elements are simultaneously present which differ in U-235 weight by more than a factor 3. If one looks at the list of the fuel elements e ASTRA-reactoith n r e wil(Tablon l) 1 enotic e thae normath t l equilibrium cor emixea (Figs i d ) cor9 . e because differenth f o e t wate controre th gap n i sl fuel elemen d standartan d fuel element. This means that mixed cores are rather the rule than the exception in the ASTRA-reactor.

4) Thermalhydraulic Analysis of Mixed Cores

The therrnalhydraulic analysis of mixed core configurations was carried out with methods outlined in several parts of the guideboo corr fo ke conversion [1]. Peak heat flu ONEt xa , peak heat flux at DNB using the Mirshak corellation and peak heat flu onset a x f floto w instability were calculate r eacfo dh fuel element type in the core using the measured pressure drop across the floa cor t wea rat f 14meo 3watea /mi d ran n inlet temperature . C ° o38 f

The measured pressure drop acros core sth e agreed fairly well with the calculated values. The pressure drop across the fuel element was calculated for different fuel element types and is shown in Fig. 1.

5) Mixed Core Configurations with Curved and Straight Type Fuel Elements

In 196e fueth 9l element ASTRA-reactoe th n i s r were changen i d several aspects mose {TablTh t. limitin1) e g aspect froe th m standpoint of operation was the transition from curved to straight fuel plate shape demandet I . a dspecifi c strategr fo y core conversion. It was no longer possible to load only one fresh fuel element in the core but it was necessary to replace at once a complete row of fuel elements. Typical core configurations in the transition phase are shown in Fig. 2. Fig. 3 indicates

27 Type of Fuel Number of Water Channel Element Fuel Plates Thickness [mm] C.F.E. 17 2.12 ST.F.E. 23 2.23 LEU-Test F.E. 20 2.44 ST.F.E.(curved) 19 2.95

0.25 2.12 2.44

0.20 "2.95

3 0.15

0.10

0.05

0.00 1234 Flow Velocity [m/s]

Fig. 1 Pressure Drop across Fuel Element for Different Fuel Element Types

the flud powean x r distributio cora n ei n configuration wite hon f fueo lw ro elements replaced e thermaTh . l hydraulic analysir fo s standard fuel elements wit plates9 1 h plate3 a contro2 , d an s l fuel element with 17 plates is given in Fig. 4. The result of e analysith s summarizei s Tablr n threi dfo 2 e core configurations,

VI/2 a core configuration before start of change VI/8 a core configuration with one row of S.T.F.E. replaced VTI/ cora 8 e configuration wito row tw f hfueo s l elements replaced includin w controgne also ltw o elements

e peaTh k hea r floo t wB fluxe DN instabilit t a s e threth ef o y core configurations analysed are indicated in Fig. 4 and are connected with a dashed line.

28 c o C o '#, CtëJ ciö o c G 0 c r, cA c Q

Core: VI/2 Core:VI/8 Core before star f replacemeno t t One row replaced Number of fuel plates in the core-446 Number of fuel plates 421 Weight of U-235 [g] in the core:3626 U-235 [g]:3700

Core: VI/13 Core:VI/19 Two rows replaced Tnree rows replaced Numbe f fueo r l plates:432 Numoer of fuel plates.387 U-235 [g]:3881 U-235 [g] :3566

too 'À À

Core: VI/24 Core: VII/43 All curved F.E. replaced Equilibrium core Numbe f fueo r l plates:421 Number of fuel plates.473 U-235 [g] 3728 U-235 (g].3733

ST.F.E. with curved plates ST.F.E. with straight plates C.F.E. *ith curved plates C.F.E. with straight plates Beryllium Reflector Element (straight) Beryllium-Irradiation Element Aluminium-Irradiation Element Graphite Reflector Element (curved)

2 FigMixe. d Core Configuration e Transitioth n i s n Period from Curved to Straight Type Fuel Elements

One can see that the peak heat flux at onset of flow instability decreased remarkably for the new standard fuel element and even more in the new control fuel element. However this effect was by compensater fa additionae th y db l numbe f fuero e b l n plateca s a s seen from the margins to DNB and flow instability in the mixed core configurations VI/8 and VI/13, unexpectedly the margins are higher in the mixed cores. After the core conversion from curved to straight type fuel element w equilibriune sa m cors ewa w equilibriuobtainedne e th n I .m corfuee th el shuffling patters nwa reversed. New fuel elements are introduced at the edge of the core.

29 10

C.F.E. C.F.E. 20 139.82 155.73 87.65 153.96 77.29 155.17 140.14 560 670 712 754 648 542 377 3.02 4.04 2.95 4.49 2.35 3.26 2.04 C.F.E. C.F.E. P.F.E. 140.90 64.84 157.09 173.59 62.76 44.10 30 675 766 857 760 680 596 3.66 2.31 5.21 5.13 1.98 1.23 S S S S S 40 250.77 257.42 258.58 259.22 258.24 650 728 733 675 563 6.58 7.33 7.48 6.85 5.69 C.F.E. C.F.E. 138.58 150.79 72.86 153.58 66.24 144.82 128.69 50 590 780 834 888 784 680 492 3.15 4.54 4.54 5.27 2.43 3.80 2.43 60 .O

= Fue S l elements with straight plates C.F.E = .Contro l fuel element U-235 weight [g] in fuel element at begin of cycle Relative average thermal neutro e fuen th flul n 1 elemenx t Thermal e fuepoweth l f elemeno r percentagn i t f reactoeo r powe) [% r

Mixe3 Fig. d Core Configuration Vl/8 with Curve Straighd dan t Type Fuel Elements Thermal Neutron Flux and Power Distribution

300

250

200

'S.!«2'95' VI/13 = 150 (2.23)

Type of Fuel Number of Water Channel 100 Elément Fuel Plates Thickness [mm] C.F.E. 17 2.12 ST.F.E. 23 2.23 ST.F.E.(curved) 19 2.95

2.0 2.5 3.0 3.5 4.0 Flow Velocity (m/s]

Fig Pea.4 k Heat Flu ON8t B (Mirshaka x DN , d Flo)an w Instabilit r Fuefo yl Element Types in the Transition Period from Curved to Straight Type Fuel Elements

30 TABL . THERMAE2 L HYDRAULIC ANALYSI MIXEF SO D CORE CONFIGURATIONS WITH CURVE STRAIGHD DAN T TYPE FUEL ELEMENTS Reactor Power: 6 MW Flow Rate: 14 m3/min. Inlet Temperature: 38° C

Core Number of Pressure Drop Flow Total Average Peak qONB lqDNB Margin to Configuration Fuel Plates across Core Velocity P.P.F. Heat Flux Heat Flux "F.I. ONB DNB F.I. 2 2 fbar] [m/s] [W/cm ] [W/cm ] fW/cm.2!» iw/ctrr f fW/cni2! curved VI/2 S.T.F.E. 385 2.64 2.25 18.68 42.04 99 222 225 2.35 5.28 5.35 curved 0.111 C.F.E. 61 2.64 3.38 18.68 63.15 99 222 225 1.57 3.52 3.56 curved S.T.F.E: 252 3.09 1.76 19.79 34.83 113 233 260 3.24 6.69 7.46 curved i/ T /o L . r . L . 54 3.09 2.62 19.79 51.85 113 233 260 2.18 4.49 5.01 VI/8 straight 0.147 S.T.F.E. 115 2.60 2.04 18.91 38.58 94 205 167 2.44 5.31 4.33 straight C.F.E. - - - - — ------curved S.T.F.E. 187 3.35 1.91 19.28 36.82 121 238 281 3.29 6.46 7.63 curved VI/13 C-F.E. 36 3.35 3.21 19.28 61.89 121 238 281 1.96 3.85 4.54 straight 0.185 S.T.F.E. 175 2.83 1.85 18.44 34.12 102 211 181 2.99 6.18 5.30 straight C.F.E. 34 2.75 1.87 18.44 34.48 102 206 167 2.96 5.97 4.84

S.T.F.E. = Standard Fuel Element

C.F.E. = Control Fuel Element to

TABL . THERMAE3 L HYDRAULIC ANALYSI MIXEF SO D CORE CONFIGURATIONS WITH CURVE STRAIGHD DAN T TYPE FUEL ELEMENTS

Reactor Power: 8 MW Flow Rate mVmin4 1 : . Inlet Temperature: 38° C

Marg in t,0 Core Number of Pressure Drop Flow Total Average Peak qONB %NB «F.I. Configuration Fuel Plates across Core Velocity P.P.F. Heat Flux Heat Flux (Mtrshak) 2 ONB ONB F.I. [bar] [m/s] [W/crn2] [W/cm2 ] iW/cm2) [W/cm2] (W/cm ] Equil ibrium core ST.F.E. 405 3.32 1.91 22.46 42.90 117 224 210 2.73 5.22 4.90 VII/43 0.227 C.F.E. 68 3.22 2.50 22.46 56.15 117 218 193 2.08 3.88 3.44

ST.F.E. 382 3.32 2.26 22.60 51.08 117 224 210 2.29 4.39 4.11 VIII/6 0.227 C.F.E. 68 3.22 1.40 22.60 31.64 117 218 193 a. 70 6.89 6.10 LEU-Test F.E. 20 3.51 2.38 22.60 53.79 125 233 236 2.32 4.33 4.39

ST.F.E. 359 3.32 1.98 22.75 45.05 117 224 210 2.60 4.97 4.66 VIII/9 0.227 C.F.E. 68 3.22 2.57 22.75 58.47 117 218 193 2.00 3.73 3.30 LEU-Test F.E. 40 3.51 2.79 22.75 63.47 125 233 236 1.97 3.67 3.72

ST.F.E Standar= . d Fuel Element C.F.E. = Control Fuel Element LEU-Test F.E. = Low Enriched Uranium Test Fuel Element Thermalhydraulic analysis was performed for a typical configuration equilibriue VII/4th f o 3 m core resulte .Th e sar summarized in Table 3.

Comparing the results with those in Table 2 one can see that this core conversion cause remarkabla d e increas corn i e e performance allowing the increase of reactor power from 6 to withouW M 8 t decreasin e safetth g y margins.

6) Mixed Core Configurations with LEU-Test Fuel Elements

Base neutronin o d thermalhydraulid an c c calculations performed in close cooperation with ANL it was decided to choose a 20 plate type LEU-fuel element (Fig U-23g a wit) 0 5 .s a 35 5h prototype test element for the ASTRA-reactor. Three LEU-test fuel elements were fabricated by NUKEM. The first LEU-test fuel element was loaded into the core in April 1982, the second in November 1982 e resultin.Th g mixed core configuratione sar shown in Fig. 7 and Fig. 8.

The thermalhydraulic analysi a standar r fo s d fuel element plate3 controa 2 d d an san wit l0 2 hfue l element wit plate7 1 h s is given in Fig. 6. The results of the analysis is summarized in r threTablfo e3 e core configurations:

VII/43 equilibriu mASTRA-reactoe th cor f o e r plate3 base2 n o ds ST.F.E. and 17 plates C.F.E. VIII/6 one LEU-test fuel element in the core. The control fuel element e corth en i shavl hig al eh burnup values. VIII/1 o LEU-testw 3 t fuel element coree th e MEU-ST.F.E ,n on i s . (4 enrichment)% 5 e fres,on h C.F.E.

The following conclusions can be drawn from the results in Tabl: 3 e

- Safety margins of LEU-test elements are by approximately 25 % lower thar HEU-ST.F.Efo n e samth e n th i positio .o t e du n higher uranium contentreducee th d dsan numbe f fuero l plates. - Safety margins of LEU-ST.F.E. are still higher than those of HEU control fuel elements.

33 Fuel Plate

0.45

i, W 20 Fuel Plates

7.61

All dimensions in cm Thouteo tw e r fuel plates hav claa e d thicknes f 0.04o s m 8c Lattice . 7.7x pitc 1 1h8.

Fig 5 .ASTRA-Reacto r LEU-Test Fuel Elemen 0 Plates(2 t ) Fuel Materia : l^CL-A l U-235 n WeighFuei ] l [g tElement 0 35 : Enrichmen : 19.5] {% t 0 Uranium density [g/cm3]: 2.84

This conclusion n agreemeni e ar s t with those drawn fron theoretical calculation d Ö.F.Z.San L s AN mad .y b e

7. Reactivity and Flux iMeasurements with LEU-Test Fuel Elements

Befor e LEU-testh e t fuel elements were loaded inte corr oth fo e power operatio a nserie f measuremento s d comparisoan s n with HEU-S.T.F.E d MEU-ST.F.Ean . enrichment% 5 (4 . ) were carried out.

34 300

VII1/6: YIII/9 _ 250 qQNB(2.44)

qDN8(2.23) ZOO

aS. 150

Type of Fuel Numbef ro Water Channel Element Fuel Plates Thickness [mm] C.F.E. 17 2.12 ST.F.E. 23 2.23 LEU-Test F.E. 20 2.44 0 4. 2.5 5 3. 3.0

Flow Velocity [m/s]

Fig 6 .Pea k Heat Flu t ON8a x (Mirshak B ,ON d an ) Flow Instability for Fuel Element Types in Mixed Core Configurations with LEU-Test F.E.

10 o Cac AI C.F.E. / 227.56 110.97 169.16 206.69 20 665 699 732 558 lurt 6.27 3.29 4.93 4.71 C.F.E. 240.70 176.52 126.95 108.36 122.00 93.59 244.62 30 579 751 899 977 805 597 459 6.05 5.30 4.41 4.04 3.78 2.16 4.70 P.F.E. LEU 271.97 156.80 157.71 143.62 215.94 349.67 622 708 812 785 566 391 7.21 4.69 5.05 4.41 5.02 5.86 C.F.E. C.F.E. 50 189.55 75.39 175.40 71.99 214.32 647 680 712 615 518 p 4.95 1.94 4.99 1.67 4.55 o o 60 Al O /Al o

LEU = LEU-test fuel element C.F.E. = Control fuel element P.F.E .= Partia l fuel element U-235 weight (gj in fuel element at begin at cycle Relative average thermal neutron flux in the fuel element Thermal fuee poweth l f elemenro percentageon i t f reactor powe] |% r

Fig 7 .Mixe d Core Configuration VIII/6 wite LEU-Teson h t Fuel Element Thermal Neutron Flux and Power Distribution

35 10 Cad

197.69 195. 30 164.06 219.71 20 698 739 780 558 5.56 6. 09 5.04 5.02

LEU C.F.E. 349.69 184.02 144. 72 106.14 122.37 83.19 242.84 30 486 776 900 976 831 594 431 7.24 5.70 5. 06 3.92 3.89 1.87 4.34 MEU P.P.E. LEU 300.49 134.56 133. 77 151.33 191.80 321.17 40 501 721 910 774 558 332 6.34 3.97 4. 69 4.57 4.30 4.47 C.F. E. 216.38 68. 98 151.62 103.89 50 192.33 617 652 687 616 545 5.45 1. 69 4.06 2.48 4.21 p o o 60

AI O M O

LE LBJ-tes= U t fuel eltanent C.F.E. Contro= l fuel element P.F.E Partia= . l fuel element U-235 wetqht |g) in fuel element at begin of cycle Kelativ average thermal neutron flux in the fuel element 'Hiormal power of the fuel elcinrnt Jn perocntntpof reactor power |%]

8 MixeFig. d Core Configuration VIII/9 witLEU-Teso htw t Fuel Elements Thermal Neutron fluPowed an x r Distribution

7.1 Reactivity Measurement

The reactivity of LEU-test fuel elements was measured in compariso o HEU-ST.F.Et n d MEU-ST.F.Ean . o differentw n i . t positions of the core, in Pos. 48 on the edge of the core e centre coreth th an n Posn i f di .o e 5 .3

MEU to HEU LEU to HEU ] [Ak/% k [Ak/k %]

Pos8 4 . + 0.01 - 0.16 Pos. 35 + 0.02 - 0.33

7.2 Flux Measurements

The thermal neutron flux was measured with copper wires in the HEU-ST.F.E., MEU-ST.F.E. and LEU-test fuel element in a position core th e t a edge ) (Pos48 . .

36 X 10 O O Cadmium Al C.F.E. 20 188.47 74.71 149.76 221.13 557 630 702 557 4.38 1.87 4.27 5.25 t.l-.t. 249.94 165.82 113.83 101.13 110.70 102.07 267.90 30 487 706 871 968 931 650 446 5.29 4.82 3.94 3.87 4.08 2.71 5.28

P.F.E. 246.18 114.83 132.37 128.91 162.40 278.97 40 502 666 832 855 667 476 5.36 3.22 4.43 4.42 4.45 5.91 1 . 1 . E . 195.49 187.92 151.63 171.63 171.50 50 .,0 625 705 785 692 599 o 5.13 5.80 4.85 5.12 5.55 Al

60 o 0 Al 0 A, O

C.F.E Contro= . l fuel element P.F.E Partia= . l fuel elemen 4 fue(1 t l plates) U-235 weighn fuei ) l (g element t begia t f cyclo n e Relative avoraye thermal neutro e fuen th flu l n i elemenx t Therma le fuepoweth lf o elemenr n percentagi t f reactoo e r powe] [% r

Fig. 9 Equilibrium Core VII/43 Thermal Neutron FluPowed an x r Distribution

From this measurement e followinth s g power peaking factors were derived:

HEU MEU LEU

Radial P.P.P. 1 .18 1 .18 1 .38 Axial P.P.P. 1 .42 1 .40 1 .38 Local P.P.F. 1 . 14 1 .20 1 .25 Total P.P.F. 1 .91 1 .98 2.38 Thermal neutron flux measurements were also performed in irradiation element spositioe closth o e t e th n f (Poso ) 48 . LEU-test fuel element, MEU-ST.F.E HEU-ST.F.Ed .an o remarkabl.N e difference coul e foundb d .

7.3 Burnu LEU-Tesf po t Fuel Elements

Marcf o d hen 198 e e ALEU-testh tth 3 t fuel elemente th n i s reactor achieved the following burnup values

ST-31 introduce Aprin i d l 1982 ...... 13.5% ST-32 introduced in November 1982 ...... 8.5 %

37 8. Conclusions

According to the experience made with the core conversion in 1969 and the recent experience with LEU-test fuel elements the following conclusions can be drawn from the standpoint of thermalhydraulics

- Operation of the ASTRA-reactor with mixed cores is possible without reducing the safety margins

- By choosing a suitable strategy for replacement of fuel element graduasa l transitio reducee th o t nd enrichment cycle is feasible.

Reference

[1] Research Reactor Core Conversiof o e nUs froe th m Highly w EnricheEnricheLo f o de dUs Uraniue th o t m Uranium Fuels Guidebook,IAEA-TECDOC-233

38 Appendix M-3

HEU-MEU MIXED CORE EXPERIMENT KUCE TH AN SI

K. KANDA, S. SHIROYA, M. HAYASHI, K. KOBAYASHI, T. SHIBATA Research Reactor Institute, Kyoto University, Osaka, Japan

Abstract In response to a request from the consultant meeting of IAEA, the HEU-MEU mixed-core experiments in the KUCA were started in April 1984. The HEU-MEU mixed-core employed in the KUCA experiment a light-water-moderate s wa s d heavyan d - water-reflected coupled-core. Several patterns of HEU-MEU mixed-cores employed in the KUCA coupled-core experiments were broadly classified into two categories. The first was called as "Separate Core" in which one cylindrical core consisted of only HEU fuel and the othe U fuele seconME rTh s .calle wa d s "Mixea d d Coren i " which each cylindrical core consisted of both HEU and MEU fuels r thesFo . e cores criticae ,th reactivite l th mas d san y worth of the control rod were measured. For "Separate Core", the effect of boron burnable-poison and the neutron flux distribution were also investigated. In both "Separate Core" and "Mixed Core" numbee ,th f fuero l plate eacn i s h cylindrical core of the coupled two cores was maintained as the same number. The imbalance of neutron importance between the two coupled core s observewa s d throug e presenth h t KUCA mixed- core experiments, since the MEU fuel plate had a slightly higher reactivity effec U fuetHE lthae plateth ne Th . reactivity worth of each control rod varied from case to case dependin e mixed-corth n o g e configuration n otheI . r words, the worth depended on the balance of neutron importance coupleo betweetw e th dn cores. However totae ,th l reactivity e controwortth f o hl rods gave approximatel same th ye value in any mixed-core configuration.

INTRODUCTION

In accordance with the joint ANL-KURRI [Argonne National Laboratory - Kyoto University Research Reactor Institute] study concernin e RERTth g R [Reduced Enrichmen r Researcfo t d Tesan ht Reactors] program e criticath , l experiments using MEU [Medium-Enriched-Uranium] fuel in the KUCA [Kyoto University Critical Assembly] were started in May 1981.*~3 The KUCA core employed in the MEU experiments was a light-water-moderated and heavy-water- reflected cylindrical core (single-core). The KUCA MEU experiments have been providing useful data with regard to the RERTR program.1*'18

e consultanTh t meetin f IAEo g A [International Atomic Energy Agency] concerning the RERTR program requested KURRI to perform a HEU-MEU mixed-core experiment. It is important to investigate the nuclear characteristics of the

39 HEU-MEU mixed-core through the critical experiments, since most of reactors cannot hardly avoid installing a mixed-core on the way to reduce the enrichment of uranium fuel for reasons of economy and reactor performance.

In response to the request from IAEA, an application for a safety review (Reactor Installation License) of the HEU-MEU mixed-core installed in the KUCA s submitte wa e Sciencth d o Technologt dan e y Agenc f Japao y n [STAJ n Juli ] y 1983. This applicatio s revisenwa Decemben i d r 198a licens d 3 s an issue ewa d in February 1984. Subsequently, an application for "Authorization before Construction s submitte "wa s approve wa d Marcn an di d h 1984. Then crite ,th - ical experiments of the HEU-MEU mixed-core were started in April 1984, as a par f "Inspectioo t n before Operation y STAr whicb " fo J a certificath s wa e issued in May 1984. The HEU-MEU mixed-core employed in the present KUCA experiment a light-water-moderate s wa s d heavy-water-reflectean d d coupled- core.

This paper provides some results of the HEU-MEU mixed-core experiments in the KUCA. The HEU-MEU experiments included the measurements of (1) criticali- ) controty(2 , worthd ro l) boro (3 , n burnable-poiso n) neutro (4 effec d nan t flux distribution.

EXPERIMENTAL

Core Configuration

Figure 1 shows a view of a heavy-water tank made of aluminum for the present KUCA mixed-core experiments. The fuel elements were assembled in a cylindrical form as shown in Fig. 2. Two cylindrical assemblies of fuel elements such as shown in Fig. 2 were installed in the heavy-water tank to for coupled-corema .

KUCe U fuels jusSincth ME Awa amoune n td si th e equa an botf o tU o t lhHE that e requirecylindricaon r fo d l assembl f fueo y l elements, patternf o s HEU-MEU coupled-cores employed in the present mixed-core experiments were broadly classified into categoriestw o e firs Th s calle.wa t s "Separata d e Core" in which one cylindrical assembly consisted of only HEU fuel and the other MEU fuel. The second was called as "Mixed Core" in which each cylindrical assembly U fuelsconsisteME d .an bot f o U d hHE e thicknes Th e heavy-wate th e minimu f th o s d man r m reflectoc 0 3 s i r thickness of the heavy-water layer between the coupled two cores is 15 cm. Each core has a cylindrical center island of light-water, and each fuel region is divide e controd th e spacint o partr th tw o lfo e y sb rods . Each inner fuel region consists of 6 fuel elements which are numbered as IN-01, IN-02 and so on for both HEU and MEU fuel elements. Each outer fuel region consists of 12 fuel elements numbered as OUT-01, OUT-02, etc. for HEU fuel elements and as EX-01, EX-02, U elements etcME r maximu.e fo Th . m numbe f fueo r l plates which e loade b a fue n n li dca elemen 5 fue1 ls i tplate r elemenn innea pe s r rfo t fuel element and 17 fuel plates per element for an outer fuel element.

A typical core configuratio s show i ne criticalit n FigTh ni e . 3 th f o y core was controlled by three rods, namely Cl, C2 and C3 rods, because all safety ) rodwerS6 s e d (S4withdrawan 5 S , o theit n r upper limi t evera t y operation. The detectors were arranged around the heavy-water tank, and the neutron sourc s locateewa d unde heavy-watee th r r tank.

40 Fig. 1. View of the Heavy-Water Tank for a Coupled-Core.

Space for Control Rods Center Island Hook

Inner Fuel Element

Side-Plate Outer Fuel Element

Supporting Grid

Fuel Plate Stopper Positioner

FigAssemblee . 2 .Vieth f wo d Fuel Elements.

It shoul D fue e noteME b dl d platedan thaU HE st cannoe mixee on b t n i d fuel element, since the dimensions of HEU and MEU fuel plates differ each other (se e1-2)d an Tabl 1 .1- e Althoug 3.8ha m fue0m l pitc employes hwa n i d the present mixed-core experiments, HEU fuel plates were originally designed for a 3.84 mm fuel pitch, while MEU fuel plates for a 3.80 mm fuel pitch. It shoul alse b d o noted tha n aluminua t m pipe (see Fig) whic2 . h separatee th s center islan f light-wateo d r froe inneth m r fuel element utilizet no s n swa i d the present experiments. In both "Separate Core" and "Mixed Core", the criticality was adjusted by the number of fuel plates inserted into the inner part of the inner fuel

41 Start-up Channels (Fission Chambers), #4 -v #6 Lin-N, Log-N and Safety Channels, Respectively (Uncompensated lonization Chambers), Cl -v C3 Control Rods, S4 * S6 Safety Rods, N Neutron Source, V Heavy-Water Reflector, OUT-01 -v OUT-1 Oute: 2 r Fuel Element FuelU HE r , sfo EX-01 -v EX-12 : Outer Fuel Elements for MEU Fuel, IN-01 *> IN-06 : Inner Fuel Elements.

Fig. 3. Typical Core Configuration of a Mixed-Core in the KUCA.

elements, and the number of fuel plates in each cylindrical assembly of the coupled-cor s maintainee samewa th ee b number o t d .

Criticality Measurement e firs th e critica ts th A ster fo pl approac e coupled-coreth f o h l al , outer fuel elements were fully loaded with 17 fuel plates. Then, the critical approach was performed by inserting fuel plates into the inner fuel elements from the outside toward inside in order. The inverse multiplication method was adopted for the critical approach. The detectors used in this measurement were three fission chambers utilized for the start-up channels, namely #1, #2 . an#3 d

For "Separate Core", the criticality measurements were performed for 4 patterns of the coupled-cores. These patterns were the cores (1) containing no boron burnable-poison [BP allt ]a ) containin (2 , innee th onlP B grn i yfue l region of the MEU side core, (3) containing BP only in the outer fuel region

42 Table 1-1. Specification of the HEU Fuel Plate.

inner fuel plate outer fuel plate plati width width curvature Uranium U-235 width width curvature Uranium U-235 no. of fuel of meat radius f meafuef o o l t radius (mm) (mm) (mm) (gr) (gr)(mm) (mm) (mm) (gr) (gr)

1 51.71 42.95 56.17 8.10 7.54 62.60 53.84 133.83 10.22 9.52 2 55.74 46.98 60.01 8.83 8.22 64.61 55.85 137.67 10.54 9.82 3 59.76 51.00 63.85 9.70 9.03 66.62 57.86 141.51 10.97 10.22 4 63.78 55.02 67.69 10.49 9.77 68.63 59.87 145.35 11.33 10.55 5 67.80 59.04 71.53 11.22 10.45 70.64 61.88 149.19 11.83 11.02 6 71.82 63.06 75.37 12.06 11.23 72.65 63.89 153.03 12.13 11.30 7 75.84 67.08 79.21 12.84 11.96 74.66 65.90 156.87 12.57 11.71 8 79.86 71.10 83.05 13.60 12.67 76.67 67.91 160.71 12.98 12.09 9 83.88 75.12 86.89 14.37 13.38 78.69 69.93 164.55 13.54 12.61 10 87.91 79.15 90.73 15.07 14.04 80.70 71.94 168.39 14.03 13.07 11 91.93 83.17 94.57 15.68 14.60 82.71 73.95 172.23 14.17 13.20 12 95.95 87.19 98.41 16.46 15.33 84.72 75.96 176.07 14.73 13.72 13 99.97 91.21 102.25 17.41 16.22 86.73 77.97 179.91 14.90 13.88 14 103.99 95.23 106.09 18.32 17.06 88.74 79.98 183.75 15.28 14.23 15 108.01 99.25 109.93 18.96 17.66 90.75 81.99 187.59 15.54 14.47 16 — — — — — . 92.76 84.00 191.43 15.91 14.82 17 — — — — — 94.77 86.01 195.27 16.42 15.29

enrichment 93.l4w% platr length = 650 mm fuel plate pitch- 3.84 mm meal lengtm m h 0 —60

Table 1-2. Specification of the MEU Fuel Plate.

inner fuel plate outer fuel plate plate no. width width curvature Uranium U-235 width width curvature Uranium U-235 of fuel of meat radius of fuel of meat radius (mm) (mm) (mm) (gr) (gr) (mm) (mm) (mm) (gr) (gr) 1 48.70 39.50 54.4 20.00 8.99 61.16 51.% 133.3 25.96 11.67 2 52.68 43.48 58.2 21.64 9.72 63.15 53.95 137.1 26.94 12.11 3 56.66 47.46 62.0 23.67 10.64 65.14 55.94 140.9 28.51 12.81 4 60.64 51.44 65.8 25.67 11.54 67.13 57.93 144.7 28.99 13.00 5 64.62 55.42 69.6 27.57 12.39 69.12 59.92 148.5 30.12 13.54 6 68.60 59.40 73.4 29.57 13.29 71.11 61.91 152.3 31.04 13.91 7 72.58 63.38 77.2 31.87 14.21 73.10 63.90 156.1 31.92 14.36 8 76.56 67.36 81.0 34.26 15.41 75.09 65.89 159.9 32.85 14.76 9 80.54 71.34 84.8 36.18 16.24 77.08 67.88 163.7 33.89 15.23 10 84.51 75.31 88.6 38.27 17.16 79.07 69.87 167.5 35.55 15.99 11 88.49 79.29 92.4 40.34 18.10 81.06 71.86 171.3 36.49 16.41 12 92.47 83.27 96.2 43.09 19.26 83.05 73.85 175.1 37.10 10.61 13 96.45 87.25 100.0 44.49 19.98 85.04 75.84 178.9 38.25 17.10 14 100.43 91.23 103.8 40.74 20.89 87.03 77.83 182.7 39.MJ I7.7lj 15 104.41 95.21 107.6 48.30 21.64 89.02 79.82 186.5 40.69 18.27 16 — — — — — 91.01 81.81 190.3 41.45 18.63 17 — — — — — 93.00 83.80 194.1 42.69 19.11 enrichment 44.87 w% plate length = 650 mm fuel plate pitch= 3.8 mm meam t m lengt 0 60 h=

43 of the MEU side core, and (4) containing BP both in the inner and outer fuel U sidME eregione coreth r "Mixe Fo f .o s d Core" e criticalitth , y measurement s performewa e pattere coupled-cordon th onlr f o fo ny e containint a P B o n g all.

l patternal r f mixed-coreFo o s s listed above e excesth , s reactivities were measured by the positive period method.

Contro Wortd lRo h

After achieving the criticality, the reactivity worths of all control and safety rods were measured by the integral count technique of the rod drop method. Three fission chamber e start-uth f o s p channel ) wer#3 s e d (#1an 2 ,# measurementse th use n i d .

The measurements were performed for all patterns of mixed-cores listed before, excluding "Separate Core" containin innee th P onl B grn i yfue l region of the MEU side core.

Boron Burnable-Poison [BP] Effect For "Separate Core" e reactivit,th s investigatedy wa effecP B f o t . From the criticality measurement 4 pattern r fo sf "Separato s e Core" described before reactivitP B e ,th y effect coul e estimatedb d .

It should be noted that special side-plates containing burnable-poison mad f naturao e l boron were prepare U fuel dME n inne A onl.r rfo y side-plate with BP contained 104 mg 10B, while an outer side-plate 117 mg 10B. It should be also noted that a fuel element consisted of two side-plates. r "SeparatFo e Core" e spatiacontaininth , lBP dependenco P n B g e th f o e effec s alswa to investigated proceduree Th . s wer s follows ea measuro t ) (1 :e the excess reactivity of the coupled-core containing no BP, (2) to substitute U fueME l f elemento e on thar s fo witmeasurto t P withouB h d excese an th e P tB s reactivity, and (3) to change the position of the substitution and to measure the excess reactivity unti spatiae th l l dependenc s obtainedwa e .

Neutron Flux Distribution

For "Separate Core", the neutron flux distribution was measured by the activation method using gold wires. From the activity measurements of gold wires with and without cadmium covers, the cadmium ratio and the thermal neutron flux were obtained.

RESULTS AND DISCUSSION

The results of criticality measurements are tabulated in Table 2 and the loading patterns of fuel plates are shown in Fig. 4. The measured results of contro d safet an ld worth ro ye tabulatear s n Tabli d . 3 Figure 5 showe e th s spatial e inneP dependenceffecB th r outee o rn i tth r f o fuee l regiof o n "Separate Core" e verticaTh . l neutron flux distributio n "Separati n e Core" containin shows i Fign P i nB . .6 o n g

44 Tabl . 2 e Critical MasExcesd an s s Reactivity.

"Separate Core" "Mixed Coupled-Core Core" No BP1 Inner BP2 Outer BP3 All BP1* (No BP)

HEU 241 250 254 258 243 Numbef ro HEU 241 250 254 258 243 Fuel Plates Total 482 500 508 516 486

HEU 3132.66 3238.66 3287.92 3335.76 3147.84

U ME 3853.6U Mas) s(g 6 3996.50 4055.74 4112.58 3886.14 23S Total 6974.74 7235.16 7343.66 7448.34 7033.98

Excess Reactivity 0.26 0.23 (ZAk/k) 0.26 0.35

Fuel Loading Pattern Fig.4(a) Fig.4(c) Fig.4(d) Fig.4(e) Fig.4(f)

MEU Core "HEU Single-Core Core" No BP Inner BP Outer BP All BP (No BP)

Number of Fuel Plates 262 286 278

235U Mass (g) 4165.74 4438.62 3542.52

Excess Reactivity 0.04 0.13 0.17

Boro1 n burnable-poiso containent [BPno s ] wa all t da .

BP was contained only in the inner fuel region of the MEU side core.

B2 P was contained only in the outer fuel region of the MEU side core. ''Bs 3 containePwa de inne th botd oute ran n h i r fueU lHE regione th f so side core.

Table 2 shows that, for "Separate Core" containing no BP, the number of fuel plates require o achievt d e criticalit s leswa y s than thar "Mixefo t d Core". This fact indicates that there was an imbalance of neutron importance betwee e coupleth n o coretw df "Separato s e Core" n otheI . r words U fue,ME l was not completely equivalent to HEU fuel in reactivity. Table 3 clearly shows the imbalance of neutron importance mentioned above e reactivitTh . y wort eacf ho h contro d varielro d cas casy eb e depending on the mixed-core configuration. For "Separate Core" containing no BP, the controU sidME ed locatea e large corro lth d ha en ri d worth thae n th tha n i t U sideHE . Ever "Mixefo n d Core" employee presenth n i td experimentse th , control rod located in the MEU dominant side core had a larger worth than that in the HEU dominant side. Such a tendency was caused by the fact that MEU a slightl fued ha l y higher reactivity effec te othe thath U fuel HE rn O . hand r "Separatfo , e Core" containin e controU th ME d locatee , ro lth BP g n i d side core was of less worth than that in the HEU side.

45 Tabl . 3 eReactivit y Wort Controf ho l Rods.

"Separate Core" "Mixed MEU Sin- Core" gle Core No BP 1 Outer BP2 All BP3 (N) oBP (No BP}1*

S4 0.13 0.34 0.49 0.22 0.73 C2 0.21 0.46 0.63 0.31 0.71 Reactivity C3 0.29 0.50 0.65 0.38 0.71 Wortf ho Cl 0.63 0.32 0.17 0.52 Control Rods5 S6 0.45 0.24 0.11 0.38 (XAk/k) S5 0.37 0.18 0.07 0.28

Total 2.07 2.04 2.12 2.08 2.16

Fuel Loading Pattern Fig.4(b) Fig.4(d) Fig.4(e) Fig.4(f)

1Boron burnable-poiso t ncontaine no IBPs wa ] t alla d . containes 2Bwa P de oute th onl n ri y fuel regioe MEth Uf sidno e core. containes 3BPwa de inne th botd oute an n ri h r fuel regione MEth U f o s side core. "*Single-core wit a h3.8 4 nan fuel pitch, other coupled-cores ("Separate Core" and "Mixed Core") with a 3.80 mm fuel pitch. 5The relative experimental error was estimated to be 2 ^ 3 %.

Tabl 3 alse o shows thae controth t d locate f ro heavyo l p dga - neae th r water coupleo betweetw e d th na large core d ha sr worth thae othersth n . This fact indicates that, the closer to the gap of heavy-water, the higher the neutron importance becomes. Suc a distributioh f neutrono n importance on s i e of the most typical characteristics in a coupled-core.

However, Tabl show3 e s thae totath t l reactivity e controwortth d f o han l safety rods gave approximatel e samth y emixed-cor y valuan n i e e configuration. Furthermore e totath , le controwortth f o h e mixed-corl th rod n i s e (coupled- core) was approximately equal to that in the MEU single-core.

Tabl show2 e s that r "Separatfo , e reactivit P Core"B e th , y e effecth f o t outer fuel region was larger than that of the inner fuel region when sideplates containin werP gB e fully loade n eaci d h region.

Figure 5 shows that, the closer to the gap of heavy-water between the two coupled substitutiocoreP B e positio th se th f no n became e largeth , r (more negative) the reactivity effect became. This fact also demonstrates that the f heavy-watero p ga e close e th highe th o , t e neutror th r n importance becomes.

Figur 6 showe s that n "Separati , e Core" e thermacontaininth , BP l o n g neutron flux in the MEU side core was higher than that in the HEU side. This fact clearly demonstrates that ME Ua slightl d fueha l y higher reactivity effect than HEU fuel.

Figur als6 e o shows thae cadmiuth t m e MEratith U n sidoi e corlowes ewa r than that in the HEU side, in other words, the neutron spectrum in the MEU side s corhardewa e r tha e HEnth tha Un i e treasoside s Th thae th .wa nt H/U ratio in MEU fuel was smaller than that in HEU fuel.

46 (a) "Separate Core" Containing No BP (employed in the (d) "Separate Core" Containing BP Only in the Outer criticality measurement). Fuel Region of the MEU Side Core.

) "Separat(b e Core" Containin P (employeB e o gN th n i d "Separat) (e e Core" Containin Innee BotP th gB r n hi measuremen e controth worth)d f o tro l . and Outer Fue U SidlME eRegione Coreth f .so

) "Separat(c e Core" Containin Innee th OnlP gB rn i y "Mixe) (f d Core" Containin. BP o gN Fuel Region of the HEU Side Core.

Note that an Arabic numeral enclosed with a circle shows a number of MEU fuel plates loaded in a fuel element, while an Arabic numeral without a circle indicates a number of HEU fuel plate«.

Fig . 4 . Fuel Loading Patterns Employe e Mixed-Corth n i d e Experiments. -0. 3

-fr-S

y -0. 4 u-0i)

-0. 5

o «

) Spatia(a l Dependenc e EffecP B th e n th i t f eo Inner Fuel Region.

-0. 2

3 . 0 - o

Ü tfl -0. 5 0) EX-12 EX-11 EX-10 EX-09 EX-08 EX-07 Position of Substitution (see Fig. 3)

) Spatia(b l Dependenc e EffecP B th e n i th t f eo Outer Fuel Region.

Fig. 5. Spatial Dependence of the BP Reactivity Effect U Sid ME f "Separat eo e ith n e Core".

ACKNOWLEDGEMENTS

The authors wish to thank all members of the KUCA for their generous help and cooperatio e experimentsth n i n . Thee verar yy gratefu . TadashDr o t li Akimoto and Dr. Masashi Tsuji's group of Hokkaido Univ., Prof. Otohiko Aizawa's grou f Musasho p i Inst f Technolo . d Profan . . Kazuhik. oDr Kudd an o Takaaki Osawa's group of Kyushu Univ. for their cooperation in the experiments. They wish to thank the members of the Technical Committee of KUCA chaired by Prof. Hiroshi Nishihara of Kyoto Univ. for their advice and support presene ith n t work.

e presenTh t wor s speciallwa k y supporte e Ministrth y b df Education o y , Science and Culture, Japan.

48 T j __ 3 0 O

O .^ 3 ————————————————— Fuel Plate ————————————————— O 30 < •^ ^ ———————————————— Fuel Meat ————————————————— ^»

o ' u O 00 o •H o Oo o0 C o o a a o A k rj* a ^ o nj À A A À a O A u u Ä U O 1 ^ 4J 2 2 0 •HA Ä o 0 ° * A k U u » 0 A A O cfl ô . •H o o! Sw - ^ O A ü 4J 0 Thermal Neutron Flux O n) X Ä • < 0 A A„ ,_( _ A o 2 t ft. A A ,_j o Oo o°*~ a O ^^ d • o k OO a A ô po j_i ^ A A A «« A • 3 i » * 0,»:MEU»ide 0) J 1 0 z t J e i i ^ A ô, • H F 1J ' J '

A 1 *«

Cadnium Ratio

0 n 0 65" Distance froLowee th m Fuere th Edgl f Plato e e (cm)

Fig. 6, Vertical Distributions of the Thermal Neutron Flux and the Cadmium Ratio in "Separate Core" Containin. BP o gN

REFERENCES

1. T. Shibata and K. Kanda, "ANL-KURRI Joint Study on the Use of Reduced Enrichment Fuel in KUHFR, Phase A Report", February 15, 1979, submitted U.S e d Japanestth o.an e Governments.

2. T. Shibata and A. Travelli, "ANL-KURRI Joint Study on the Use of Reduced Enrichment Fue KUHFRn i l , Status Repor n Phas o t, Decembe B" e , 198012 r , submitte U.S e d Japanesth . an o t d e Governments. . Shibat T Travelli. A d . aan 3 , "ANL-KURR ReduceIf o Join e Us tde Studth n o y Enrichment Fue n KUHFRi l , Joint Repor n Phaso t, Marc , B" e1983 31 h , submitted to the U.S. and Japanese Governments.

A. K. Kanda, S. Shiroya, M. Hayashi, Y. Nakagome and T. Shibata, "KUCA Critical Experiments Using MEU Fuel", Paper IAEA-SR-77/30 presented at the Semina n Researco r h Reactor Operatio d Usean n , Julich, Federal Republic of Germany, September 14-18, 1981.

49 5. M. Hayashi and S. Shiroya, "Few-Group Constants for thé HEU and MEU Cores e KUCA"th in , Annu. Rep. Res. Reactor Inst. Kyoto Univ., vol.14, pp.153-164 (1981).

6. K. Kanda, S. Shiroya, M. Hayashi, K. Kobayashi, Y. Nakagome and T. Shibata, "KUCA Critical Experiments Using Medium Enriched Uranium Fuel", ibid., vol.15, pp.1-14 (1982). 7. S. Shiroya, H. Fukui, Y. Senda, M. Hayashi and K. Kobayashi, "Measurement Neutrof so n Flux Distribution a Mediu n i s m Enriched Uranium Core", ibid., vol.15, pp.141-150 (1982).

Kanda. K Nakagom. ,. Y 8 Hayashi. M d ean , "Licensing Procedure d Safetsan y Criteria for Core Conversion in Japan, Safety Review of KUCA Conversion", in Proceedings of the International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reac- tors, Argonne, Illinois, USA, November 12-14, 1980, ANL/RERTR/TM-3, CONF-801144, pp.36-50 (August 1983).

9. K. Kanda, "KUCA Critical Experiment Program Using HEU and MEU Fuel", ibid., pp.567-578 (August 1983).

Kanda. K . Hayashi. ,M 10 Shiroya. ,S Kobayashi. ,K . Fukui,H Mishira. ,K d an a T. Shibata, "KUCA Critical Experiments Using MEU Fuel (II)", in Proceed- e Internationath ing f o s l Meetin n Researco g d Tesan ht Reactor Core Conversion FuelsU LE o st , froArgonneU mHE , Illinois, USA, November 8-10, 1982, ANL/RERTR/TM-4, CONF-821155, pp.426-448 (September 1983).

. Shiroya11S . Hayashi. ,M Kanda. ,K Shibata. ,T . E Woodruf. L . J . ,W d an f Matos, "Analysis of the KUCA MEU Experiments Using the ANL Code System", ibid., pp.449-461 (September 1983).

Tsuchihashi. K Mor. 12T d .an i Analysin ,"A JAERKUCe f sU o Coreth AME I y sb SRAC Code System", ibid., pp.462-471 (September 1983).

. Fukui13H . Mishima. ,K Shiroya. ,S Hayashi. ,M . Nishihara H Kand. ,K d an a , "Experimenta le Voi th Studd n o Reactivity y Coefficien e KUCA"th n ,i t Annu. Rep. Res. Reactor Inst. Kyoto Univ., vol.16, pp.1-16 (1983).

14. S. Shiroya, M. Hayashi, K. Kanda, T. Shibata, W. L. Woodruff and J. E. Matos, ^"Analysi KUCe U Experimentth ME A f o s L Cods AN Usin ee System"th g , ibid., vol.16, pp.17-33 (1983).

15. S. Shiroya, M. Hayashi, K. Kanda, and T. Shibata, "Analysis on the KUCA MEU Experiments (II), Boron Burnable-Poison Effect", in Proceedings of the International Meetin n Reduceo g d Enrichmen r Researcfo t d Tesan ht Reactors, Tokai-mura, Japan, October 24-27, 1983, JAERI-M 84-073, pp.369-376 (May 1984).

16. M. Hayashi, S. Shiroya, K. Kanda, and T. Shibata, "The Calculation of the MEU-HEU Coupled Core in the KUCA", ibid., pp.377-387 (May 1984).

Kanda. 17K . Mishima. ,K Shiroya. ,S Hayashi. ,M Fukui. . Nishihara,H H d ,an , "Experimental Study on the Void Reactivity Coefficient in the KUCA", ibid., pp.388-398 (May 1984).

. SendaJ8Y . Shiroya. ,S Hayashi. ,M Shibata. Kand. T , K d aan , "Analysin so the Void Reactivity Measurement in the KUCA", ibid., pp.399-408 (May 1984).

50 Appendi4 xM-

THE TRANSITION PHASE OF THE WHOLE-CORE DEMONSTRATIO RIDGK OA EE RESEARCTH T NA H REACTOR

R.W. HOBBS Oak Ridge National Laboratory*, Oak Ridge, Tennessee M.M. BRETSCHER, RJ. CORNELLA, J.L. SNELGROVE Argonne National Laboratory, Argonne, Illinois United State f Americso a

Abstract The transition from operation of the Oak Ridge Research Reactor with high-enrichment uranium (HEU) fuel to operation with low-enrichment uranium (LEU) fuel is nearing completion. The systematics of the replacement of thU fue eHE fueU ldiscussede LE lwitar e hth e resultcore Th .th e f physso - ics measurements that have been conducted during the transition phase are described.

INTRODUCTION

k RidgOa ee ResearcTh h Reactor (ORR bees )ha n selectea hoss r da fo t full-core demonstration of the newly developed USi2 low-enrichment uranium

(LEU) fuel. LEU fuel elements were first introduce3 d into the reactor in Januar thif yo s year. Currently, onlhigh-enrichmeno tw y t uranium (HEU) control rod elements remain in the operating core, and it is expected that operation with a full LEU core will begin in late December. The transitio s beenha n accomplishe graduaa y b d la U fue LE n phase-i li e th f no manner consistent with normal operatio ORR e methoe th Th . d schef no an d - dule followed during this transition are presented in this paper. During the phase-in period, various core physics measurements have been conducted to provide data to validate the core neutronics calculations being performe Argonny b d e National Laboratory (ANL) additionn I . , measuremento t s determine fuel element burnup and to verify that required safety margins were met were also performed. Details of these experiments and a summary e resultoth f e presentedsar .

BRIEF DESCRIPTION OF THE OAK RIDGE RESEARCH REACTOR

e operatinTh g configuratio introductioe R prioth OR o t re th f f o nno U elementLE cor7 y 9x shows ean i s e matriFigurn Th ni beryllius . i x1 e m reflecte n threo d e side wated an s r reflecte fourthe th n containt o dI . s U oxidHE e7 2 elementcontrox si elementd d ro lsan s with fueled followers.

The fuel element is a box type containing 19 curved Al-clad plates and a total of 285 gU oa235 fs U0a when new. The fueled control rod followers

are constructed with3 15 fuel plates and contain a total of 167 g 235U each.

* Operated by Martin Marietta Energy Systems, Inc., for the US Department of Energy under contract DE-ACO5-84OR21400.

51 The LEU fuel elements and control rods are geometrically identical to the

HEU elements and control rods. Only the fuel meat section has been

changed. LEU fuel elements contain a total of 340 gU oeac235 f h as

USi when new, and the shim rods contain 20U0 a235 gs U3Si each when 2 2 new3 . Experiment locatee ar s corn i d e positions B-l, B-9, C-3, C-7, E-3, E-5, E-7, F-l, and F-9. During the phase-in period, the ORR operating configuration has slowly changed in response to the removal of experiments and various safety reviews e currenTh . t operating configuration, 177-Ds ,i show Figurn ni . e2 e operatinTh gr configuration cyclfo W M e 0 3 tim t a e s simila thoso t r e shows approximateli 2 n Figuren i d an e 1 sth yf o thre d en e e weeksth t A . three-week operation, all fuel elements are removed from the core and

HSST

LEGEND: F=FUEL fl BE F F F F F F F BE SR = SHIt1 ROD BE=BERYLLIUn B EU BE F SR F SR F BE EU EU=EUROPIUn IR=IRRIDIUH C BE F riFE F F F MFE F BE MFE=ttRGNETIC FUSION D EU F F SR F SR F F EU ENERGY HSST=HEflVY E BE F «FED F IR F IR F BE SECTION STEEL TECHNOLOGY F EU BE F SR F SR F BE EU HFED=niNIPLflTE EXP. G BE BE BE BE BE BE BE BE BE

123456789

FIG. 1. ORR core lattice configuration prior to phase-in of LEU.

LEGEND: F=FUEL ELEMENT R DE BE BE F F F BE BE DE SR = SHIf1 ROD =E 8ERYLLIUrB 1 B BE BE F SR BE SR F BE BE flLUf1INU= L fl n D DUHrlE= Y FUEL. EL C BE F riFE F F F MFE F BE t1FE = ttflGNETIC FUSION ENERGY D BE F F SR F SR F F BE HFED=MINIPLflTE EXPERIMENT E BE F HFED F F F flL F BE

F BE BE F SR F SR F BE BE

G DE BE BE BE BE BE BE BE DE 123456789

FIG. 2. ORR core configuration 177-D.

52 stored in the reactor pool to allow for Xe decay. Typically, three to four of these elements have achieved 50% burnup and are declared spent. The reacto s theri n refueled with thre fouo w elementt e rne d fuesan l elements of various 235U content froinventore th m irradiatef o y d fuel e storeth n i d controx si pool elementd e ro lTh . s locate positionn i d s B-4, B-6, D-4, D-6, F-4, and F-6 follow a different refueling scheme, since they are not returne pooe th l o aftet d r each operating cycle e schem Th ., tha eis t after approximately three month operationf so w unirradiatene o ,tw d control rods are introduced into positions D-4 and D-6. The control rods from positions 6 whicD- hd an hav 4 e D- been irradiate threr dfo e month movee sar posio t d - tions B-4 and B-6, and the control rods from positions B-4 and B-6 are moved to positions F-4 and F-6. The control rods from positions F-4 and F-6 which have been in the reactor for approximately nine months of operation and have achieved approximately 70% burnup are declared spent and are moved to pool storage.

PHASE-IN OF LEU FUEL

e objectiv Th transitioe th f o e n perioreplaco t s i d e spen U fueHE t l elements and control rods with fresh LEU elements and control rods and to irradiate this fue o obtailt inventorn na f elementyo controd san l rods wit235e th Uh burnups ORRe typicath .r Thilfo s phase-i bees nha n accomplished by installing three to four unirradiated LEU elements rather thathree th nfouo t e r unirradiate elementU dHE s normally introduced during each refueling. Thereafter, the LEU fuel is used in the same manner as the U fuelHE , i.e., after each cycl irradiationf eo fueU placeLE s li e ,n th i d pool storage for one operating cycle to allow for Xe decay and then returned to the operating core. Thus, an inventory of LEU elements with various 235U burnup becomes available and an all-LEU core is established. U fue HE s wit lA e elementsth h U contro LE phase-ie e ,th th l f rodno s ha s been accomplished by introducing two LEU control rods rather than the two unirradiate U controHE d l rod approximatelt sa y three-month intervald san shuffling the rods as previously described.

The sequence of fuel management followed during the phase-in period is show . 4 ExaminatioFiguren ni d an 3 s thesf no e figures showU sLE thao tw t cores were being developed; the first core consisted of elements irradiated e secon d cycleth od d de an s consisteth n i f elemento d s irradiatee th n i d even cycles e firsTh . t cycl operationf o e , 174-DE, used thre U eleLE e - elementU HE fue7 totaa 2 4 r ment2 lf lo fo sd element an s d operatean s d from January 7, 1986, to February 1, 1986. All six control rods in this core were HEU. After irradiation in cycle 1, these three LEU elements were stored in the pool during the operation of cycle 2 and then returned to the core along with four fresh LEU elements to give a total of seven LEU elements in cycle 3 (175-A). Cycle 5 used the seven LEU elements irradiated in cycle 3, plus three additional fresh LEU elements for a total of teU elementsnLE . Similar comment remainine sth holr cyclesd fo d god e Th . even-numbered cycle e configuresar cycled dod similae s th containin o rt g the same number of LEU elements and follow the same fuel management scheme just described e firsTh . tcontroU paiLE f ro l rods were loaded into core positions D-4 and D-6 at the start of the cycle 7 (176-B) so that two of the six control rods were LEU. Two additional control rods were added in cycle 11 (177-B) to bring the number of LEU control rods to 4 out of 6. It is expected thae fina th o tcontro ltw l rods wil loadee lb d intcore t oth a e the beginning of cycle 15 (178-A) scheduled to start in mid-December. This will be the first all-LEU core and will mark the end of the transition demonstratione phasth f o e .

53 tl 13 • 5 t7 1/7-2/1 3/3-3/26 4/16-5/3 6/27-7/21 3/27 0/6 7/27 0/6 10/27 0/6 14/27 2/6 174-D+5 E 175-A 175-C 176-B

•2 • 4 16 2/4-2/20 3/26-4/16 5/12-5/31 3/27 0/6 6 7/20/ 7 10/27 0/6 174-F 175-B 176-A

2/1-2/4 2/22-3/3 5/3-5/12 5/31-6/27 END CYCLE II END CYCL2 1 E END CYCLE »5 UNSCHEDULED SHUTDOWN

fleasureaentt on Cor» 174-FX Approach to critical Measursments on Core 176-AX1 Reactor shutdowe du n U74-F similas i X o 17S-At r ) measurenents <176-AX s similai t o 176-Bt r ) R seismiOR o t c review 7 LEU Fuel Eléments Cores: U FueLE l4 1 Elements and subsequent siphon- 0 LEU Shift Rods HEU-1 U ShiLE m2 Rods break Modification HEU-2 . Flu1 x Napping 174-FX LEU-1 1. Flux Mapping 176-AXl . Shi 1 d CalibrationRo « s LEU-2 176-B 2. £•*«/! - unsuccessful 2. ß.««/l - unsuccessful £•*«/!. 2 , 176-d Ban Shift Rod Calibrations 3. Shi« Rod Calibrations 176-BX2 (All HEU) 174-FX 176-AXl . Baftft3 a Heating 4. HSST Reactivity Measurements - Measurement unsuccessful

DATA FORMAT Cycle number Date U elements/totaLE l element U shiLE m s rods/total shim rods Core configuration ______

FIG. 3. Schedule for transition to LEU (cycles 1 through 7). 19 tit 113 tlS B/12-9/1 9/26-10/15 11/4-11/20 12/15-1/4 6 17/22/ 5 6 21/24/ 3 24/26 4/ 4 24/24 6/6 176-D 177-B 177-D 178-A ftLU LE L 18 #10 «12 «14 7/21-8/11 9/11-9/26 10/16-11/3 11/21-12/7 14/27 2/6 6 17/22/ 5 6 21/24/ 5 24/24 4/6 176-C 177-fl 177-C 177-E

9/2-9/11 12/8-12/15

END CYCL9 * E END CYCLE I 14

Measurements on Corp 177-AX1 Measurements on Cycle »15 (all LEU) (177-AXl is similar to 177-B) Core 178-A 21 LEU Fuel Elements 24 LEU Fuel Elements U Shi*LE m Rods ShiU LE m 6 Rods

. Flu1 x napping 177-flXl 1. Flux Mapping

2. Ö...W1 177-AX2 2. ß.**/l . Shi3 d CalibrationRo « s 3. Shim Rod Calibrations 177 AX1 . Gamn4 a Heating . Gamm4 a Heating Measurements Measurenents- unsuccesful

DATA FORMAT Cycle number Dato LEU elements/total elements LEU shit rods/total shim rods Core configuration______

. SchedulFIG4 . transitior efo (cycle U LE o nthrougst 8 h 15). FIG. 5. Flux monitor holder.

Figures 3 and 4 show that there were periods during which the reactor was shut down for scheduled maintenance and quarterly inspections. During" varioue thesth , e s9) period cord an e , sphysic6 (end, 5 cyclef , so s2 , 1 s measurements listed were mad provido t e e datneutronicr afo s code validation and to obtain information necessary to verify that the required safety margins were maintained. These measurements are discussed in the following sections.

FLUX MEASUREMENTS

Determination thermae th f so l neutron flux distributio varioue th n i ns core configurations have been made using 0.020-inch diameter Co-V wire (2.0 wt % Co in V) and measuring the induced 60Co activity. The Co-V wire is inserted axially into the water channel between two fuel plates using the aluminum holder shown in Figure 5. This holder accommodates two full- length Co-V wire d allowan s wiree s th eacexteno st f ho d approximatele on y inch above and below the active fuel region. Each fuel element in the core is monitored with one flux monitor assembly inserted into the ninth water channel counting from the concave side of the element. Additional monitors have been inserted into channels thre d fifteeelemente ean th n ni s wite th h

56 ÂVG VALUE OVER ACTIVE FUEL LENGTH

50 2 '4 6 B 10 12 14 16 IB 20 22 DISTANC WIRF E O FRON INCHEI EP MTO S

FIG. 6. Typical axial flux profile.

highest power density. An underwater camera is used to ensure both the locatio d alignmenn an monitor e th elementse f th o t n i s . Control rods, whic difficule har accesso t t , have been monitore insertiny db Co-e th gV wire tubl A int en oa (0.080-inch diameter) whic thes hi n inserted inte th o water channel betwee fuee th nl fueplatee th l n followeri s controe th f o sl rods. The precise location and alignment of these monitors are uncertain. A similar arrangemen l tub(A te with Co-V wire bees )ha n use monitoo t d r fluxes in several Be reflector pieces that surround the fuel.

Co-e th V f wireo givea l n si Al n test configuratio e irradiatenar d simultaneously for six hours at approximately 300 kW. After irradiation, the wire removee sar counted an dcomputer-controllea n o d d wire scanning system. In this system, the Co-V wire is wrapped around a circular disk. e dis Th indexes i k determino t d e axial position e alon wiree th th gd ,an computer control steppina s g motor which move dise predeterminen th si k d increments behind a tungsten collimator and initiates the predetermined counting interval.

A discriminator window is set on the output signal from the Nal crystal detector so that only the twCoo60 photopeaks are recorded. e measureTh d 60Co counting rates versus axial position forwardee sar o t d ANL to be compared with calculated values of the relative thermal flux distributions in the fuel elements.1 For most of the core, the agreement between measured and calculated values is quite good. A typical plot of these dat shows i a Figurn ni . 6 Note reflecto e th e r flux peak neae rth the bottom of the fuel. In this analysis, only relative values (counting rates) have been determined. Absolute measurements require an accurate determination of the reactor power during the Irradiation. This is dif- ficul obtaio t n with sufficien w powet lo accurac re th leve t a yl maintained durin e irradiatioth g flue th x f monitorso n .

For the Co-V wires closest to the fuel element center in each element, an axially averaged value of the Co counting rate may be determined. The power produce fuea y lb d elemen assumes i t e relate e producb th o t do f t do t 60

57 Table 1. Calculated safety margins

Core Configuration Core Element power Safety Fuel elements Control rods position ___(MW)_____ margin LEU HEU LEU HEU Meas. Calc.

174-C 0 27 0 6 C-6 1.30 1.29 2.12

174-FX 7 20 0 6 D-7 1.27 1.24 2.52

176-AX1 4 13 2 4 E-4 1.31 1.36 2.13

177-AX1 1 4 4 2 E-4 1.40 1.39 1.87

this average counting rate time 235e Uth s weight (adjuste burnupr fo d n )i the fuel element. 2 The element power can then be determined by normalizing the sum of the above products to the reactor power. As mentioned earlier, severaelemente th f lo s producing highest power contained additional flux wires . The measured neutron flux gradients , determined by comparing the activit multiple th f o y e wiresingla n i s e fuel element usee detero ,ar t d - mine power density peaking factors withi elementse th n . From these data, y determinma e maximue on th e m thermal heat fluxes during operatioe th d nan safety margin, which is the ratio of the critical heat flux to the maximum heat flux at the limiting conditions of operation (43.5 MW and 14,100 gpm flow) methodologe Th . y use gives i d Referencn ni . e Result3 e th r fo s limiting elemen foue th n eacri t f coreho s measure dat o t e gived ear n ni Tabl . Not1 e e that r cor,fo e 177-AX1 increase ,th fuen ei l element power d reductioan safetn ni y resulte margith e decreasinf nso ar e corth ge size core U th fuele LE f I siz f .d beeo eha t attributee n no us e e th ar o d t dan maintaine elements7 2 t a estimateds i t ,i d thae safetth t y margin woule b d 2.10, which is comparable to the margin in the all-HEU core 174-C. The minimum allowe d 14,10 an 43.t coolanm da W M 05gp 6 margi 1. ts i nflow . It shoul e noteb d d thae fluth tx measurement d safetan s y margin analyses are made prior to 30-MW operation of a new configuration. As an f cyclo configuratio, d 1 eexample en e th t ,a n 174-F n s loaderu Xwa d an d e flu th w powex a lo r tmeasurements fo r . Core 174-FX contained seveU LE n elements (four new LEU elements and the three irradiated in core 174-DE) and was similar to cycle 3 (175-A). After the measurement, this core was removed and core configuration 174-F installed. The analysis of the flux measurements was made and approval obtained prior to operation of core 175-A. Similar comments hold for cores 176-AX and 176-B and cores 177-AX1 and 177-B.

Measurement prompe th f tso neutron decay constan = (3/£ a t, whers i 3 e the effective delayed neutron fraction and I the prompt neutron lifetime, have been made using noise analysis techniques.1* Signals from two fission

58 chambers located on opposite sides of the core periphery and near the core midplane are processed by a Fourier analyzer to obtain the cross-power spectral density (CPSD) as a function of frequency. The CPSD is then fitted, using least-squares techniques, to obtain the break frequency, Fjj, where a = 2irF]j. The measurements are made at very low reactor power (approximately 3 kW). Measurements on small, cold, clean water-reflected cores (Figures 9 and 10) consisting of all fresh HEU elements (285 g) and shim rods (167 g) and all fresh LEU elements (340 g) and control rods (200 g) show the prompt neutron decay constant to be about 14% larger in the LEU core than in the HEU core. This agrees very well with the ANL calculations. These measurements were made witfissioe th h n chamber operatin pulse th en i gmode .

Measurement operatinn i a f o s g cores have been mad n configurationo e s 176-BX2 (all HEU) fue7 f 2 o lf ,U o element LE U 176-BX 2 LE d 4 san (1 1 6 shi fue5 f 2 o m lf U rodso element LE U d 177-AX4 )LE an d 1 an s(2 1 6 shim rods). Result beet sye nhavt obtainedno e o conducT . t these measurements, special amplifiers were designed to allow the fission chamber to operate in the current mode. Additional measurements of a are planned U coresLE d . an n ful o U lHE

GAMMA SCANNIN FUEF O G L ELEMENTS After each cycl operationf eo fuee ,th l element e removesar d froe th m core for one cycle to allow for Xe decay. During this inter-cycle time, each fuel element irradiated in the previous operating cycle is scanned along its centerline to detect y-rays from the fission products. The experimental apparatus is shown schematically in Figure 7. The element is placed horizontally, convex side up, in a tray located 16 feet under water. The tray translate e elementh s t beloy tubdr ew a 1/2 inc diameten i h coms i d -ran puter controlled to stop at predetermined counting locations for a specified time. The dry tube (which acts as a collimator) extends up through the 16 feet of pool water to the bottom of a lead collimator with a 1/16-inch- diameter hole Ge-LA . i crystal detecto centeres i r d oveleae th rd colli- mato connected ran Nucleaa o t d r Data 6600 data gathering systemM IB n A . s interconnectei C P Nucleae th o t dr Data Systemresulte th y-peaka d f ,so an fitting routine for each axial location scanned are stored on a separate floppy disk for each element. For each isotope, the measured counting rate for a given energy line s relatei s irradiatioit o t d n histor d decaan y yequation a tim e y b eth f no form:

Activit [F« yD (X, tirrad^ tdecay F^D+ (A , tirrad) , (1 tdecay) 1 1 2 2 2 2

+ ...... FnDn (A, tirradn, tdecayn)] where, Activit measure« y d counting rate

Fn = the average fission rate during operating cycle n,

n D productioe =th decad an n y facto cyclr fo r, en irradiatioe tirrath = d n time duringd cyclan , en

decae tdecath y= ytim e measured froshutdowe mth f cyclno en to counting time. 59 FIG. 7. Fuel element gamma scanning.

For a long-lived isotope such as Cs, the measured counting rate is proportional to the total number of fissions 137 . For a short-lived isotope, all contributions from irradiation cycles other than the most recent are negligible measuree th d ,an d count rat proportionas i e e fissioth o t ln rate e lasth t n i cycl f operationo e . Intermediate-lived isotope usee b o t dn ca s determine the fission rate during the last cycle of irradiation (F ) if valuee otheth f F^Do sr ^fo cycle f irradiatioo s knowne ar n .

Initially, it was hoped that Cs could be detected and used to integrate the fissions to determin e137 the burnup of each element. Due to the abundance of short-lived fission products and the small elapsed time be-

tween irradiatio d countin an nelemene th f go t (typicall o thret o e tw yday s after irradiation)Cs, 137 cannot be unambiguously detected. The predomi- nant isotopes remaining after approximatel dayo tw ys deca 11+0e ar yBa , llt0La, 132I, 95Zr, "Mo, and 95Nb. The 1596 KeV line of 1J+0La, which decays wit effectivn ha e half-lif bees ha f 12.n o e , beed 8 n selected an d sorted from the data to use in the analyses due to its high yield and energy determininy B . elemene gth t fission rat eacn ei h cycle 235e ,th U burnup can be estimated, and in addition, confirming information on the core power distributio obtainede b y nma .

60 Tabl givee2 comparisosa calculatef no measured dan d fuel element power r corsfo e 176- determines Aa d frogamme th m a scanninge th dat n ai following manner:

Table 2. Comparison of calculated and measured fuel element powers core 176-A

Element power Core position (MW) C/Ea Calc. Meas.

A-2 0.641 0.701 0.914 A-3 0.767 0.846 0.906 A-4 0.924 1.031 0.897 A-5 1.074 1.124 0.955 A-6 0.928 1.029 0.902 A-7 0.794 0.823 0.965 A-8 0.605 0.659 0.918 B-3 0.925 1.007 0.919 B-5 0.987 1.032 0.957 B-7 0.867 0.888 0.977 C-2 1.227 1.216 1.009 C-4 1.122 1.112 1.009 C-5 1.109 1.241 0.894 C-6 1.372 1.411 0.972 C-8 1.179 1.158 1.018 D-2 0.792 0.827 0.958 D-3 1.318 1.274 1.034 D-5 1.064 1.061 1.003 D-7 1.233 1.152 1.070 D-8 0.790 0.765 1.033 E-2 1.035 0.983 1.052 E-4 1.241 1.177 1.055 E-6 1.234 1.137 1.086 E-8 1.067 0.958 1.114 F-3 0.758 0.676 1.122 F-5 0.867 0.740 1.171 F-7 0.766 0.661 1.158

aCalculated to experimental values.

. Spectra1 l data were acquire axia2 1 t la d positions along each fuel element. e countinTh . 2 e g159 th peaV rateacr 6f Ke fo ko e h axial location along the fuel elements is sorted from the spectral file. These counting rates are then corrected (reduced) for the residual La remaining from previous cycles of irradiation. The corrected counting rate (proportional to the La formed in the last cycle of irradiation only s the)i nshutdowf o corrected en f o ne decar th fo do t y the last cycle.

61 cUn3 in

in a.

6 8 JO 12 AXIAL POSITION

FIG. 8. Typical data from fuel element gamma scanning.

3. The trapezoidal rule was then used to integrate the area under the curve generated from the 12 points, and an axially averaged value of the ^'•'La activity at shutdown was determined. This value is then proportional to the fuel element power during the last cycle.

e absolutTh . 4 e fuel element powe s therwa n determine normaliziny b d g the axially averaged counting rates to the total power generated by the elements as determined by averaging the beginning-of-cycle and end-of-cycle element powers calculate ANLy b d .

A typical plot of the counting rate data from one element is shown in Figure 8. Note that calculated to experimental (C/E) ratios are high in the F-row core positions and low in the A-row core positions. This is discussed in others papers given at this meeting.1'^ To determine an element's 25g turnup, the accumulated MWd of irradiation on the element are determined from the measured element power and operating time. This value (MWd) is then multiplied by a burnup constant (g235U/MWd givo )e t 235j th e j burnup. Notburnue e valuth th e f o ep constant (g235U/MWd) varies (decreases) with the total MWd due to the contribution of Pu to the fission rate. A comparison of 235U burnup determined by this diffusioL AN y methob d nan dcalculation gives i s elemenr Tabln ni fo 3 e t elemente C-023th r Fo s. irradiate dateo t d , calculate measured an d d 235U burnup normalle ar s y agree withio t d o percenttw n .

CONTROL ROD WORTHS e wort controe Th th f eac ho f o hl rod measures i s d usin positive th g e period method. Tabl list4 e measure e th s d integra worthd ro lthrer fo s e core configurations measured. To accomplish this measurement one of the six control rods is fully inserted and the remaining five are ganged to

62 Tabl . Fuee3 l element C-023 irradiation history

Full power 235u 235u days of Calc. Exp. Calc. Exp. mass mass Cycle operation power power MWd MWd calc. exp. C/Ea (d) (g) (g)

174-D 12.86 1.16 1.12 14.89 14.36 321 322 1.00

174-E 10.62 1.11 1.13 26.72 26.38 307 307 1.00

175-A 18.52 1.01 0.94 45.42 43.72 285 286 1.00

175-C 17.39 1.36 1.37 69.14 67.56 256 257 1.00

176-B 21.86 0.99 0.83 90.72 85.60 231 235 0.98 176-D 19.45 0.72 0.81 104.63 101.25 214 217 0.99

Calculated to experimental mass.

Tabl . Result4 e f controso calibrationd ro l s

Rod worth %AK/K Controd ro l Cycle 174-Ca Cycle 176-Bb Cycle 177-AX1C position 12-15-85 06-16-86 09-08-86 235u ^ « 7235 g

F-4 1.254 1.936 1.909

F-6 1.289 1.896 1.786

B-4 4.181 5.624 3.045

B-6 4.301 5.241 3.271

D-4 3.984 6.852 4.687

D-6 4.682 6.118 4.749

TOTAL 19.691 27.667 19.447

All-HEU core with 27 fuel elements and six control rods. a b This cor fue7 2 e U (D- lf contain LE o D-6d element4U o an )LE tw 4 d 1 s san x controosi f l rods. c This core contains 21 LEU of 25 fuel elements and four LEU (D-4, D-6, B-4, and B-6) of six control rods.

63 maintain reactor criticality d 'on-seatro e Th s the.i 1 n withdrawna measure dresultine amounth d an t g positive period determined. From tabulated values of period versus reactivity addition, the differential rod worth at this location is determined. The rod 'on-seat' is then withdrawn e other 2 incheth d s san gange criticalt a d e differentiaTh . l worts hi measure t thia d s location measuremene th d ,an s repeatei t 2-incn i d h increments until the upper limit of rod travel is reached. Integral rod worths are obtained from the differential worths. This procedure is repeated for each of the remaining five rods. In addition to these data, tabulations of the control rod position versus time into the operating cycle are provided to ANL for comparsion with fuel cycle calculations. Currently, calculated and measured values of control rod worths are not in good agreement5 and further investigation into this is planned.

COLD CLEAN CRITICAL CORE MEASUREMENTS

At the end of cycle 2, the reactor was shut down to conduct measurement coldn o s , clean, critical coresU LE consistin d l fresan al U f hHE o g fuel element d shian sm rods. These measurements were intende provido t d e the simplest benchmark data for the ANL calculations (i.e., no fission products, well-known fuel distribution in-corw fe d e,an experiments).

Four different critical configurations were established using standard approach-to-critical procedures (based on subcritical multiplication). These four configuration loadine th d gsan sequenc e showar e Figure n ni s 9 through 12. The two Magnetic Fusion Energy Experiments (MFE) in core wer 7 t removeC- positioneno d an d 3 froe coreC- s th r mfea possiblf fo so r e experimentse damagth o t e .

The following measurements were conducted on the water-reflected cores LEU-1 and HEU-1.

Approac• criticao t h l • Shim rod calibrations Reactivit• y wort 7 MFE-6f o hC- n Ji • Core flux mapping by activation of Au wires

n LEGEND: F=FUEL ELEHENT B SR=SHin ROD r1F MFICNETI= E C F FUSION c HFE MFE B ENERGP EX Y D F SR F SR F B 1 1 1 B LORDING SEQUENCE E F F F F F S GIVEI Y B N S. 1 1 1 3 NUflBER IN LOUER F SR F SR F RIGHT HflND CORNER F 4 1 1 1 5 CRITICRLITY RTTflINED G N FIFTO H FUEL RDOITION IN F-7 123456789

FIG. 9. Water reflected core HEU-1.

64 LEGEND: R F=FUEL ELEMENT SR=SHIMROD B MFE=HRGNETIC FUSION F F F ENERGP EX Y C HFE 8 8 8 nFE D F SR F SR F LORDING SEQUENCE 3 1 1 1 6 IS GIVEN BY E F F F F F NUMBER IN LOUER 2 1 1 1 5 RIGHT HRND CORNER F SR F SR F F M 1 1 1 7 CRITICRLITY RTTRINED ON SIXTH FUEL G ROOITIO7 D- N I N 123456789

FIG. 10. Water reflected core LEU-1.

LEGEND: R F=FUEL ELEMENT SR=SHIHROD B BERYl.LIU= E B n MFE'MRGNETIC BE BE BE BE BE C nFE HFE FUSION 6 6 .,£ „._£. 6 ENERGY EXP BE BE SR F SR BE BE D 6 C 1 1 1 6 6 LORDING SEQUENCE BE BE F F F BE BE E 6 4 1 1 1 3 6 IS GIVEY B N NUMBE N LOWEI R R BE BE SR F SR BE BE F B 2 1 1 1 8 6 RIGHT HflND CORNER G BE BE BE BE BE BE BE CRITICRLITY RTTRINED 6 1 1 1 1 1 6 ON FIFTH BE RDDITION IN D3 123456789

FIG. 11. Beryllium reflected core LEU-2.

LEGEND: R F-FUEL ELEMENT SR=SHIMROD B BE=BERYLLIUM MFE=MflGNETIC BE BE BE BE BE C 6 HFE g g 6 nFE 6 FUSION ENERGP EX Y BE BE SR F SR BE BE D 6 6 1 1 1 6 LORDING SEQUENCE BE BE F F F BE BE E 6 4 1 1 1 5 6 IS GIVEY B N NUMBE N LOWEI R R BE BE SR F SR BE BE F 6 2 1 1 1 3 6 RIGHT HflND CORNER BE BE BE BE BE BE BE CRITICRLITY RTTRINED G 6 1 1 1 1 1 6 ON FIFTH BE flDDITION IN E-7 123456789

. FIGBerylliu12 . m reflected core HEU-2.

65 Measurements conducte Be-reflectee th n o d d core include:

Approac• criticao t h l • Shim rod calibration Reactivit• y wort MFE-6f o h J experiment

Prio assemblino t r g these cores, calculations wero t eL madAN y b e determine when a given configuration would achieve criticality and to determin additionae th e l amoun f fueo tberylliur lo m require attaio t d n sufficient excess reactivity for control rod calibrations. During the course of measurements, these calculations were shown to be quite accurate, Compariso f calculateo n d measurean d d quantitie e givear s Referencn ni . 5 e

CONCLUSIONS

The transition phase of the Whole-Core Demonstration at the ORR is pro- ceeding smoothly changeo N . normao t s l operating procedures have been required. With the exception of the control rod calibrations, the agreement between measurements and calculations "is good. Safety margins greater than that required have been maintained throughout this perio mixef o d d core operation. These margin e comparablsar thoso et e that have existen i d HEU cores operated prior to the beginning of this demonstration.

REFERENCES

. CorneliaJ . R Bretscher. 1M . . ,M . HobbsW . R ,, Compariso Calculatef no d and Measured Irradiated Wire Data in Support of the ORR Whole-Core Demonstration of LEU Fuel (to be published).

2. T. P. Hamrick and J. H. Swanks, The Oak Ridge Research Reactor - A Functional Description, ORNL-4169, Vol Appendi, 1 . C x (September 1968).

3. J. L. Snelgrove, unpublished information 1985.

Mihalczo. T . Rega. J E d . ,nan G "Promp. 4 t Neutron Decay Constane th r fo t Ridgk Oa e Research Reacto 235% r Ut witw Enriche0 2 h d e Fuel"b o (t , published).

5. M. M. Bretscher, "Analytical Support for the ORR Whole-Core LEU U3Si2-Al Fuel Demonstration" (to be published).

66 FULL CORES

Appendix M-5

OPERATIONAL IMPACT ENRICHMENW LO F SO T URANIUM FUEL CONVERSION ON THE FORD NUCLEAR REACTOR

R.R. BURN Ford Nuclear Reactor/Universit f Michiganyo , Ann Arbor, Michigan, United State f Americso a

Abstract Ae RERTs th par Rf o tProgram whole-cor,a e demonstration usinU LE g fuel Fore begath d n Nucleai n r Reacto Decemben i r r 1981. Numerous core performance measurements were made on a full HEU core, full U U coresfuele measurementHE LE Th d mixed .an ,an dU coreLE sf o s included control rod worths, temperature and void coefficients, full core flux maps, and in-core and ex-core spectral mesurements. Overall o significan,n t operational impacts resulted U fuelfroLE o m.t R conversiofroU FN HE me th f o n

INTRODUCTION The University of Michigan Department of Nuclear Engineering and the Michigan Memorial-Phoenix Project have been engaged in a cooperative effort with Argonne National Laboratory to test and analyze low enrichment fuel in the Ford Nuclear Reactor (FNR) e efforTh .begus twa 1979n ni e th ,pars a f o t Reduced Enrichment Researc Tesd han t Reactor (RERTR) Program demonstrato ,t e on a whole-core basis the feasibility of enrichment reduction from 93% to below 20% in MTR-type fuel designs. firse enrichmenTh w tlo t uranium (LEU) cor loaded s ean wa R d FN Inte th o criticallt achieves ywa Decemben o d 1981, r8 . Critical loadin s followegwa d by a period of about six weeks of low power testing and 3 months of high power testing during which control rod worths, full core flux maps, and in-core and ex-core spectral measurements were made.

FUEL DESIGN U fue s designeLE l wa e simila e hige b Th th o h t d o enrichmenrt t uranium (HEU) fuel, hence all existing fuel handling equipment and procedures can be used with the LEU fuel. The similarity in the fuel design greatly simplified the HEU to LEU fuel conversion. originae Th U aluminulHE m allotwenty-onr fo y R fueFN e le th use yearn i d s had an overall plate thickness of 0.060 in. (0.020 in. clad-0.020 In. meat-0.20 in. clad). In subsequent HEU aluminide fuel, utilized after 1978, overall plate thickness was reduced to 0.050 in. by reducing clad thickness to 0.015 in. (0.015 in. clad-0.020 in. meat-0.015 in. clad). The enrichment reduction from 19.5o t accomplishes % %wa 93 increasiny b d 238e Ugth loadino t gg fro0 m8. 691 g per element and by increasing the 23% loading from 140.6 g to 167.3 g

67 r elemenpe overcomo t t resultane th e t reactivity loss cause y resonancb d e absorption. The extra uranium loading was accommodated by increasing the meat weight fraction from 14.2 42.0%o %t , increasin uraniue th g m density from 6 g/cc4 g/c1. 0. increasind o ,ct an g meat thicknes e resulTh 0.03o t s-. 0in tant overall plate thickness was restored to 0.060 in. (0.015 in. clad-0.30 in. meat-0.01 . clad) in 5watee Th p thicknes.ga r numbed an s platef ro eler pe s- mene identicaar t o thost l fuen i e l that had-r beefo R n FN utilize e th n i d over twenty years. Thermal hydraulic performance of LEU fuel was not an issue in obtaining a license for its use.

CONVERSION SCHEDULE

Initiall singla y U elemeneLE s installe twa reactoe th n teso i ds t r it t integrity singlA . e elemen s followetwa whola y b de core, critical loading experiment. Transitions between LEU, HEU, and mixed cores occurred over the next three years accordancn ,i e witfollowine th h g schedules numerous ,a s core performance measurements were completed.

Octobe , 19822 r 1 Installed firs elementU tLE . Performed integrity test. December 8, 1981 Loaded initial whole LEU core. Performed critical experiment. May 10, 1982 Restored full HEU core. Remeasure U parametersHE d . Decembe 198, r6 2 Began phased transitio coreU LE o .nt

June 7, 1983 Loaded whole LEU core. Verified LEU core measurements. September 30, 1983 Restored mixed LEU-HEU core. Completed HEU burnup. October 11, 1984 Achieved full LEU core. Removed last HEU element from core.

CORE PERFORMANCE CHARACTERISTICS

It is difficult to precisely compare operating parameter measurements for U coresLE d .an CorU HE e sizes (numbe elementsf ro configurationd )an s (element arrangement in the core grid) vary between cores. The amount of fuel burnup of specific elements, particularly control elements within which control rode ar s inserted, causes significant variation parameten si r measurements generaln I . , HEU measurements for a large, equilibrium core were compared to LEU measurements for a smaller core with almost fresh, uniform burnup fuel during the threconversionU LE o et yeaU o typicarHE Tw . l core configuration showe sar n in Figure 1. The equilibrium HEU core contained 37 elements. The LEU core contained 29 elements for some measurements and 33 elements for others. Rod positions within the cores are shown on the figure. Thermal flux levels were measured usin self-powerea g d rhodium neutron detector.

68 North North

Heavy Water Tank Heavy Water Tank

A C A C Rod Rod Rod Rod

B C ontro B ontro Rod Rod Rod Rod

) Equilibriu(a Elemen7 3 m CorU tHE e (b) 29 Element LEU Core

Figure 1. Ford Nuclear Reactor Core Configurations

Thermal Flux (n/cm2/sec x 10~13) HEU Core LEU Core Center Element 2.15 1.70 Peripheral Element 1.49 1.25 In-Core Water Gap 3.56 3.64 Heavy Water Tank 2.41 2.64

fuelo tw Wer se eth identical woule ,on d have expecte highea d r in-core thermal flux with LEU fuel because of the smaller core volume. In fact, the reverse was true, indicating that the LEU core has a harder in-core flux and highea r percentag powef o e r generation results from fast fissione Th . harder flux increases fast leakage thermae th d l,an flu largn xi e in-core water gap d externa san core th e o actuallt l y increases e wit U th fue hLE s la fast leakage neutrons thermalize. Since thermal neutron irradiations are generally conducted adjacent to the core, the use of LEU fuel may actually enhance a facility's irradiation capabilities. Rod worths were measured in a 37 element, equilibrium HEU core; a 26 element, fresh LEU core; and the first 37 element, equilibrium LEU core.

Equilibriu U coremHE , JulElement7 3 198, y1 0 s

A Rod 2.2422 %Ak/k Excess - 2.98 %Ak/k B Rod 2.1354 C Rod 2.3794 Controd lRo 0.3251 TOTAL 7.0821

69 Initial LEU Core With Excess Added, December 10, 1981 26 Elements

d Ro A 2.2198 %Ak/k Excess 2.95 %Ak/k B Rod 2.3203 C Rod 2.2833 Control Rod 0.3822 TOTAL 7.2056

Equilibriu U CoremLE , Octobe 198, r11 4 37 Elements

A Rod 2.3211 %Ak/k Excess =3.20 %Ak/k B Rod 0399 C Rod 2565 Controd lRo 0.3718 TOTAL 6.9893

e totafirste Th wortd th ro l f ,ho smal U cor s somewhalLE ewa t greater than the HEU total. The equilibrium LEU core total rod worth was slightly less than the HEU total. From an operational viewpoint, the differences, causee whicb y y corb dhma e size, overall core burnup pattern d particu,an - larl specifie th y c burnu f contropo l elements that surround control rods, o consequencen arf o e .

Power defect negative th , e reactivity inserte y increasinb d g power from zero to 2 MW, was measured experimentally for one LEU and two HEU cores.

HEU Core September 1979 -0.21 %Ak/k May 1982 -0.31 %Ak/k LEU Core July 1983 -0.25 %Ak/k

Temperature coefficien measures i t establishiny b d g steady state power wite reactoth h r under automatic controcontroe th f o ll rod. Cooling systems are secured d reacto,an r power heat 50,00e th s 0 gallon poo whicn i lcore hth e is immersed e reactivitTh . y inserte automatiy b d c adjustmen controe th f to l rod is in direct reaction to the reactivity loss produced by the increase in pool temperature.

HEU Core March 1981 -7. 5io~x 3

LEU Core July 1983 -7. x IQ-9 3 %Ak/k

e temperaturTh e coefficient coul expectee b d increaso t d magnitudn i e e becaus increasef eo d resonance absorptio slowins a n g down length increases at higher temperatures change th t e, bu observe s insignificantdwa .

Void coefficien approximates i t insertiny b d a thig n aluminum bladf o e known volume into various core locations relativele ,th crosw ylo s section aluminum producin voia g displaciny b d g water. Void coefficien extremels i t y sensitive to core location and ranges from a maximum magnitude of -1.2 %Ak/k/ core centee %voi+0.o th th t e f n 2o ri d %Ak/k/%voi overmoderaten i d d locations. n averagA e valu +1.f eo 0 %Ak/k/%voi FNRe th d imperceptabluses i d,an t a d e differences were seen between HEU and LEU fuel.

70 CONCLUSIONS

No significant operational impacts have resulted from conversion of the FNR to LEU fuel. Thermal flux in the core has decreased slightly; thermal leakage flux has increased. Rod worth, temperature coefficient, and void coefficient have changed imperceptibly. Impressions from the operators are that power defec s increaseha t d slightl thad yan t fuel lifetim s increasedeha . fulls i R y FN convertee U fuelTh LE o .t d Remaining bein s i fres U ghHE shippek RidgOa o et d National Laboratory ;beins i spen U g tHE shippe e th o t d Savannah River Plant. The FNR license and technical specifications are scheduled for renewa 1985n li U fue beinHE s licensee li .f Th o g e eliminateddus .

71 Appendi6 xM-

FULL COR FUEU ELME DEMONSTRATIO JMTE TH R N I

M. SAITO . NAGAOKAY , . SHIMAKAWAS , , . NAKAYAMAF . OYAMADAR , . OKAMOTY , O Oarai Research Establishment, Japan Atomic Energy Research Establishment, Oarai-machi, Ibaraki-ken, Japan

Abstract The joint ANL-JAERI program for the RERTR was started in January e fina1980Th .l goa thif o l s progra achievo t s i m e the full core conversion to LEU in the JMTR.

In 1980, the LEU Fuel with suicide had not been qualified yet. In this situation, the MEU fuel was selected for the first stageU fue s ME confirmed integritwa lan e , th f do y throug followine hth g three steps; (1) hydraulic test, ) critica(2 l experiment JMTRe th Cn si (Japa n Materials Testing Reactor Critical Facility), and ) irradiatio(3 JMTRe n th tes .n i t

Augusn I t 1986fule th ,l cor fueU eME l demonstration test had been successfully completed.

INTRODUCTION

The JMTR is a light water moderated and cooled 50 MW tank type reactor usin typR gET e fuel elements therd mane ,an ear y irradiation facilities such as in-core capsules, hydraulic rabbit tubes, in-pile loops and a shroud facility as shown in Table 1, and the specification of the MEU fuel is shown in Table 2. U fuels r convertina ME Fo e guidline p e ,th u e desig th th t o se t f ngo s follows: e numbeth ) f fuea o r l elements e corloadeth en i e dincreaseshoulb t no d d orden i maintaio t r fase nth t neutron flux level, dimentione th ) b fuef so l elements shoul unchangede b d , c) the fuel elements should be currently qualified up to 1.6 gU/cm3 den- sityd ,an d) the U-235 content per fuel element should be sufficient to allow operation wite samth he cycle characteristic currene th U cores a sHE t . e JMT d beeth U fueRha ME n n i l e proceedeth f o e n i d us e prograe Th th f o m compliance with the following steps; (1) hydraulic test, (2) critical experiments in the JMTRC, and (3) irradiation tests of two MEU fuel elements in the JMTR. e fulTh l core U fues conversioME permittewa le th y Japanesb o dt n e governmen tabove baseth en o dexperimenta l results.

73 Table 1 Characteristics of JMTR

Type Tank type Power thermaW M 0 5 l Moderator/ coolant material H20 pressure 14 kg/cm2G temperature 47 °C (Inlet), 55°C (Outlet) coolant velocity 10 m/s Reflector Be Fuel material UMX-A1 enrichment 45% loading 8 kg of 235-U type ModifieR ET d Controd ro l f rodH s5 with 5 fuel followers Neutron flux (xlO1"4 n/cm2»s),(max.) fas1 MeV(> t ) thermal fuel region 4 4 reflector region 1 4 Power density (ave.) 0 kW/49 £ Experimental facility Capsules Hydraulic rabbit tubes Loops Shroud

Tabl 2 e Specificatio U fuelME e s th f no

Standard fuel Fuel follower

Material UA1X - Al l UA1A - X Enrichment (%) 45 45 Meat U-Density (g/cm3) 1.6 1.6 Dimension (mm) 0.5 x 62W x 760L 0.5 x SOW x 750L Clad Thickness (mm) 0.385 0.385 Fuel Dimension (mm x )780 W L71 1.2x 7 1.27 x 60W x 770L plate No. of plates 16 Fuel 19 U-235 content (g) 310 205 element Dimension (mm) 76 * 76 * 1200L 64 x 64 x 890L

The full core MEU fuel demonstration test began in July 1986, and had been successfully completed in August 1986.

Development U fues beeLE ha le nth startenexe f so e th t th s a stedr fo p full cor U fuele LE wite . hth

74 Hydraulic Test

Hydraulic test using hydraulic test facility were planned to (1) measure the coolant velocity distribution between fuel plates, (2) confirm the strength of the standard fuel elements by exposing them to hydraulic forces developed with up to 140 percent design flow, (3) determine the critical velocity, and (4) confirm withstandin f fueo g l followe n droi r p tests.

The results of hydraulic tests are shown in Table 3.

Table 3 Results of Hydraulic Test

Test Item Typ Fuef o e l Element Test Result Coolant velocity distribution Standar& d Good equalization measurement fuel Follower Confirmation of the strength Standard Strong enough undee th r against 140% of design 6 hours test velocit m/s0 (1 y) Determinatio f criticano l Standard Enough withstanding velocity agains 0 m/set2 c Calculated critical velocity (maximum velocity m/se5 1 n disregar ; (i c f o d tese oth ft facility) the strength of fuel core) ; 18 m/sec (on the assumption that fue same l th ecor s eha strengt e claddinth s a h g material) Drop test Fuel Follower Strong enough under the following test conditions o 100 time drop-test at 140% (14 m/sec) of the design velocity tim0 2 o e drop-test a t 160% (16 m/sec) of the design velocity o 20 time drop-test at 180% (18 m/sec) of the design velocity

The critical velocit s estimate wa ye approximatel b o n disret di s m/ - 5 1 y e strengtth gar f o df fue o h l resultmeate e criticath Th .f o s l velocity test showed thae MEth t U d fuewithstandinha l g against hydraulic t forcea f o s e maximu whics th leas m/ s i h0 m 2 t velocit f thayo t facility.

The fuel follower was drop-tested up to 100 times at 140 percente (14 m/s) of average velocity. Further drop tests were conducted up to 20 times each at 160 and 180 percent of average velocity (16 m/s and 18 m/s, respectively).

After every test, coolant channel gaps of fuel elements were measured and no channel gap change was observed.

75 Critical Experiments

The purposes of the experiments are to obtain nuclear characteristics and to validate neutronics calculation performed by SRAC code system.

Critical experiments in the JMTRC are as follows; (1) critical mass (2) excess reactivity, (3) contro worthsd ro l , (4) flux distribution, ) ß/Z(5 , ) shu(6 t down margind an , (7) void coefficient.

The results of main critical experiments are shwon in Tables 4-7 and e standarth e dJMTR th cor s n i Cshowi e e validit n e i FigurTh nth . f 1 o ey neutronic calculations were confirmed through these experiments.

Irradiation Test

Irradiation tests of two MEU fuel elements were carried out in the JMTR and post irradiation examinations were conducted in the hot laboratory. Items of PIE's are as follows; (1) sipping test, (2) measurement of swelling (3) oxide layer thickness measurementsd an , (4) dimensional measurements.

The result f PIE'o s s includin e fueth gl f o elementbur p u n e showar s n i n . 9 Measure d Tablan 8 e d oxide laye e calculateth r d thicknesy8 man ~ d5 s swa value was 21 pm. No swelling was observed at 0.45 x 10 fission/cm of burn up. The other examinations showed to be in good condition.

Tabl 4 eJMTR C Fuel Element Loading

Kind of Plater pe s Uranium 2350 Element Element Density, g/cm3 Contentg , MEU Standard fuel A 19 1.6 310 B 19 1.4 280 C 19 1.3 250 Fuel follower 16 1.6 205 HEU Standard fuel A 19 0.7 279 B 19 0.6 237 C 19 0.5 195 Fuel follower 16 0.7 195

76 Tabl 5 e Calculated Exess Reactivity, Contro Worthsd Ro l , Shut-Down Margin d Voian d Coefficient, Comparing with Measured One r JMTRfo s C

U ME HEU Ap Calculated Ap Measured Calculated Measured (Cal-Meas) (Cal-Meas) Exess reactivity 11.2 11.5 0.3 10.0 10.6 +0.6 XAk/k Critical mass 5077.4 5108 +30.6 4746.8 4741 -5.8 , U-23g 5 Control rod worths %Ak/k SH-& 2 SH-1 11.3 11.7 +0.6 11.7 12.5 +0.8 SA-1 3.1 2.9 -0.2 3.2 3.1 -0.1 SA-2 5.9 6.0 -fO.l 6.3 6.4 +0.1 SA-3 3.4 3.2 -0.2 3.4 3.3 -0.1 Shut-down margin 16.4 18.2 +1.8 %Ak/k 14.0 15.3 +1.3 Void coefficient -0.012 -0.013 -0.001 -0.012 -0.013 -0.001 %Ak/k/void-%

Table 6 Calculated Kinetics Parameters, Effective Delayed-Neutron Fraction d Prompt-Neutroßean ff n Life Time £,p, Comparing with Measured Ones for JMTRC

U ME H E U Measured Calculated C/M Measured Calculated C/M 0eff/*pc se > 111 125 1.13 103 118 1.15 ßeff - 0.00766 - - 0.00766 - tp, ysec - 61.1 - - 64.8 -

Table 7 Calculated Thermal Flux Changes by Core Conversion from HE Uo MEt U Fuel Comparing with Measured Ones for JMTRC

Measured Calculated Thermal neutron flux Fuel region -8 ~ -12% - 8-13~ % Be reflector region -1 ~ -3% % -3 ~ 0 Fast neutron flux Fuel and Be reflector region - % -2 + 2~

77 K L K J I H G F B C DE N 0 P 0 1 9 PM 2

3 P) p)

4 GANMA- RAY SHI LD

5 (JR-l) O o O O 6 0 syi l o Ic O B B O c 0 0 7 Qj[9 IA IA IA IB ^ ___ iJL Ü O 8 'o ISH-l SA-2 SH-2 O B B IA Ü y 9 I IB U Ü A O IA U O SA^-3 O 10 O le Ü B B Ü le 11 0 0 o £R-)

12 O O, O

14 © BF3

Remarks

Fuel Element [il Control Rod with Follower Fuel •Kin f Elemeno d t O Nock-u f Capsulo p) 30 « ( e Be Element

AI Element O Hock-u f Capsul) po 36 U e (l irradiation hole)

AI Element Q Nock-u f Capsul) o p 40 » I e {4 irradiation holes)

Hock-up of Loops and Hydraulic Rabbits

1 FigJMTR. C Standard Core

Full Core MEU Fuel Demonstration

e fulTh l core demonstration witU fues performeME hwa l d Jul , 1988 y 6 through August 2, 1986 with satisfactory results.

e importanTh t characteristic o converU t fues ME lo t thav e been already measure mentiones a d d above. Following items were performe fule th l n coro d e wit U fuelshME . (1) excess reactivity, stucd ro ke marginon ) (2 , (3) shut down margin, ) temperatur(4 e coefficient, (5) characteristics of burn up, (6) radioiodine concentration in primary cooling water, and (7) sipping test.

The main results are shown in Table 10 and Figures 2 and 3. The demonstration core configuration wit U fuelhME s showsi Figurn ni . e4

78 Tabl 8 e Swellin f MEo g U Test Fuel Elements

Irradiation Uranium Ave. Thickness Swelling Element Position Density Fission Density Change (%Av/v) (g/cm3) (fis/cm3, Cal.) (mm) Meas. Cal. x 100.4251 H-8 SM-1 1.6 (28.4) U . %B 0.0 0.0 2.8 0.44 x 1021 SM-2 J-8 1.6 0.0 0.0 2.8 (28.1) U . %B

Tabl 9 e Oxide Layer Thicknesf o s MEU Test Fuel Elements

Oxide Layer Thickness Element (mm) Meas. Cal.

SM-1 0.005 ~ 0.008 0.021

SM-2 0.00 0.005~ 8 0.021

Table 10 Comparison of Measurements and Calculation Result f MEo s U Full Core Demonstration

Measured Calculated Excess reactivity %Ak/k 10.6 11.8 Shut-down margin %Ak/k 21.5 20.5

Temperature coefficient 4 1 4k/k/°C at 30 °C 1.02 x KT 1.25 x IQ" *

Concluding Remarks

The full core demonstration with MEU fuebeed ha nl successfully completed. Some neutronics data such as excess reactivity, shut down margin, temperature coefficient c weret , e measure compared an d d with neutronics calculatione Th . result satisfactorye ar s .

During the reactor operation, fission products leakage was carefully checked by the primary coolant analysis. Sipping tests were also performed to

79 I0f

10° f 7

-• •" S™ïl

i er 10' «j E 7

° 00 0° o

—• ——LEU(mini-pk)te) irradiation test —MEUISMI.2) irradiation test 10 10* ————— HEU ——————————

5 7 " TA 3 7 F 7 7 "68 "6"6 T 67 " "69" 70 " 1984- ———— 1985 ————— ———— 1986 (yeor) Operotion cycle number

Fig 2 . Radioiodine Concentratio n Primari n y Cooling Water of JMTR

CD & MEÜ Core

100 200 300 400 500 600 Accumulated Power (MWD)

Fig 3 .CorU CharacteristicME ed an U HE r Burfo f s o p nu

80 Fuel Elracnt Loopi «od Brdraillc UbUt.

l M ei««t

Fig 4 . MEU Fuel Demonstration Core Configuration

e MEchecth Ue resultss f failed fuei kth wa s l A o fission . , n products were observed.

n additionI e brieth , f e reviehydrauli th s mad n wwa o e e c th test d an s critical experiment e JMTRCth f n whico i s, h results were presentee th n i d previous meeting.

REFERENCES

1. F. Nakayama, Y. Itabashi, H. Kanekawa, R. Oyamada, M. Saito and T. Nagamatsuya, JAERI, "Progres f o reduces d enrichment progra n i mJMTR" , Int'l. Mtg RERTRn o . , Tokai Japan

2. Y. Nagaoka, K. Takeda, S. Shimakawa, S. Koike and R. Oyamada, JAERI, Critical experiment f coreo JMTRU s sME C (II)", Int'l. Mtg n RERTR.o , Petten Netherland

81 Appendix N TRANSPORTATION, SPENT FUEL STORAGE, AND REPROCESSING

Abstract

Information is provided on transportation of fresh and spent fuel elements, spent fuel storage, and reprocessing in the U.S.

A variet f transporo y t container r fresfo s h fued an l spent fuee describear l d along with certain contractual, transportation, reprocessing batch size d economicaan , l considerations. Examples are provided of specific fresh fuel transport regulations in the FRG (as of August 1982) d administrativan e procedure n Japai sr transporfo n f o t fresh fuel elements.

Methods and results of criticality analyses for storage of HEU ,U fuel LE e presentedMEUar sd an , . Results include fissile loading, fuel element geometryd an , storage rack geometry considerations.

U.S. Federal Register Notices (as of 30 December 1987) on DOE's "Receip d Financiaan t l Settlement Provisionr fo s Nuclear Research Reactor Fuels e provided"ar . Appendix N-l TRANSPORTATIO FRESF NO H FUEL ELEMENTS

Appendix N-l.l

TRANSPORTATIO FUER LMT F ELEMENTNO S WITHI FEDERAE NTH L REPUBLI GERMANF CO Y

TRANSNUKLEAR GmbH Hanau, Federal Republi f Germanco y

Abstract Since April 1982, a regulation has been introduced in Germany that require u (inludinP s d transportan U g HE MT f o Rs fuel elementso t ) be performe a specia y b d l safety vehicle called "SIFA" summarA . y of transport regulation f Auguso s a s t 198 s i 2provide d along with data on fresh fuel shipping containers and the SIFA safety vehicle.

1. Introduction

The transport of MTR fuel elements can generally be performed by road, by rail by sea or by air. Transnuklear GmbH transported MTR fuel elements through Europe almost exclusively by road. e pasth (TransnukleaN T tn I r GmbH s executeha ) a seried f o s transports of MTR fuel elements by road to almost every research reactor operator. In addition, TN has also transported MTR fuel elements in a combined road/air/road transport to the United States of America.

Since April 1982, in Germany a regulation has been introduced, that transports of HEU and Pu can only be performed by a special safety vehicle called SIFA (Si cherheitsfahr- zeug). MTR-transport e alsar so subjec o thit t s regulation. TN has in the mean time carried out three MTR fuel element- transports wite SIFA th n hparticularI . , transportR MT f o s fuel elements have been made to KfA (Kernforschungsanlage) in Julich and to Sweden.

. 2 Transport regulations

The transportatio f o radioactivn e materiale th y roab sn i d Federal Republic of Germany is subject to the GGVS regulations. These regulations are closely based on the ADR, which cover the international road transport of radioactive materi als.

85 For the transportation of rad. materials by air, the JATA regulation sr transportatioapplyfo e d IMCth an , Oa se y b n regulat s applion y. These regulations are based on the IAEA-regulations for the Safe Transpor f Radioactivo t e Materials (1973).

. 3 Historw regulationne f o y s (see poin 1 abovet r physicafo ) l protection in Germany (SMC - _S_afety J^easu rement _C_atalogue) In September s 1976informewa N C T woul,SM e db w d thane a t prepared by the BMI (Federal Ministry of the Interior). In July 1977 the announced SMC draft was a subject for dis- cussioe "Associatioth n i n f Nucleao n r Fuel Cycle Companies". TN prepared objections to the draft -- without success. The SMC was then put into effect by the (ierman authority (PTB), responsible for, among other things, the transport of radioactive material n Decembei , r s forwarde1977wa t bu ,o t d TN only in April 1978. The transition period will occur as fol 1ows: 15 months transition status starting from the time f notificationo , i.e., valid until October 1980; thereafter assuming full responsibilit C catalogueSM e th f o y. TN received a study in which a "SIFA" (Safety vehicle) was specified by the SMC. e studth d y an raise C SM d e questionTh se techniwhicth n i h - cal sector can only be resolved by specialized companies. After initial thorough research it was established that the final specified regulations could not be reached in the 15 months indicatedo t e ;du this wa s - administrative measures - technical measures relatin e vehicleth o t g , - the fact that a specific communication system did not exist Startin e summeth n f 1977i o gr ,N informeT d companies with nuclear activities, for example, Alkem, Nukem and nuclear research centers, of the measures in this catalogue. Ia letten r date 6 Apri1 d l a B informe 1980PT d an e th N ,T d large number of other companies that, according to an order from BMI dated 3 April 1980, the Safety Measures Catalogue, edition dated December 1977, and that the repeatedly extended transition status will thereby be terminated.

Decision on construction of the "SIFA" Following publicatioI orderBM e N decideT ,th f o o connt d - struct the vehicle required. Design and construction started in June, 1980. Construction took about 18 months. The vehicle came into forc n Aprili e , 1982.

86 4. Execution of transports since the SMK 77 came into force e SIFth - vehiclAf o e Us e According to the SMC the transport of highly enriched uranium (more than 250 g U-235/transport) or plutonium (more than 1Ü g Pu/transportU )e carrie b hav o t t witee SIFAou dth h - vehi cle. This armored vehicle whic s protectei h f o d e againsus e th t firearms and explosives, is equipped with a communication system with redundancy. The armed crew is security cleared and specially trained. In addition the transport is kept under surveillanc a hig y b he security control center wita h radio control system coverin e wholth g e Federal Republif o c Germany r Pu/Fo . U - quantitie s exceedin 2 kg/transporg r o t / U-235/transporkg 5 n additionaa t l escort vehicle wita h communication n armesystea d dan m crew mus e providedb t .

5. Shipping container e shipmenFoth r f MTR-elementso t 2 type, f packagino s s i g available, namel e e modeMTR-bir th th yK 161 U ld 2an d cage. Both packaging B unde PT e approve e ar rs th y b d approval no. D/4031/F (Rev. 4) D/4033/AF (Rev. 2)

5.1 Description of MTR-bird cage The MTR-bird cage is a type A packaging, gross weight 306 kg. It consists of a steel frame bird cage in special form made of tubes having the external dimensions of 770 x 770 x 1440 mm. In the bird cage itself there is a steel inner container with dimension: 3b8 x 350 x 1020 mm supporte n centrai d l positio a stee y b nl tube container. I fe y inne spacth theran n s ri e i e containe r besidee th s element it must be filled out in appropriate way so that the elements will be locked and cannot move. e authorizeTh d contene containeth f o t s maxi r .9 MTR - fuel elements not having more than 200 gr U-235 each. In case fissilth e e conten f elemento t s differeni s t from the above mentioned quantity, the allowable quantity is 1,8 kg U-235/packaging. Authorized enrichment is up to 93,5 %, authorized activity max. 150 mCi. Maximum number of container/shipment is 6.

5.2 Description of UK 1612 e containeTh K 161a U typr s A i 2epackaging . Gross weight is 282 kg. It consists of a rectangular steel box (Design . 1612no ) wite exterioth h r dimensionx m m f abouo s0 53 t 762 mm x 20155 mm, with 8 inner transport positions.

87 top view without lid

Main Data;

Gross Weight: 300 daN Authorization: Type A

CONTAINER FOR MTR-FUEL ELEMENTS Assembly Drawing

Authorized conten r packagingpe t :

a) max. 6 nonirradiated tubic MTR fuel elements (active length: no less than 72 cm, per element con- f o n U-23taininmax o % % 3 5 9 . 9 2 enricheg d uranium with max. 403 g U-235 in form of U/Al-Alloy (altogether max. 2,41 g U-235))k 8 .

) b max .7 nonirradiate d rectangula fueR lMT r elements with 19 plates each, or MTR control elements with 16 plates each (type GE), containing per element up to 559,44 g on about 93 % of U-235 enriched uranium with max. 520 g U-235 in form of U/Al-Alloy (altogether max. 3,64 kg U-235).

c) max. 8 nonirradiated tubic MTR fuel elements (active length o lesn :s than 60,9 , activcm 6 e diameter: max. 10,41 cm) in form of U/Al-Alloy, containing n abouo % tf 0 o U-233 r maxelemen- % pe o t en 51 .8 p u t riched uranium with maxg U-23 0 .21 5 (altogether max. 1,68 kg U-235)

88 Main Data:

Weight: (calculated) empty 162 kg g k 2 ful28 l

BNFL-CONTAINER DESIGN 1612 Assembly Drawing

d) max. 8 nonirradiated MTR fuel elements (active length: between 38,1 and 63,5 cm, active surface cross-section: max. 79,03 cm2) incorporating uranium-meta U-23% 4 9 5 f maxo l . enrichment n i for,f o m U/Al-Alloy sandwiched in an Al-sheet with a U/A1 mass ratio from max. 0,14. Containing per element max. 300 g U-235 (altogether max. 2,4 kg U-235) e) max. 8 nonirradiated tubic or rectangular MTR fuel elements (active length: between 35,56 and 63,5 cm, active surface-cross-section: max. 87,1 cm2) incorporating uranium-meta U-23% 4 9 5 maxf o l . enrichment n i for, f o m U/Al-Alloy with a U/A1 mass ratio from max. 0,14. Containin r elemenpe g t max g U-23 Ü .3U 5 (altogether max. 2,4 kg U-235)

89 Package classification

authorized content mentioned under: a) M c) d) e)

fissile class: II III II II II

transport index: 12,5 50 33 33 12,b

max. no. of container/shipment: 15 15

A) Registration No. - motor tractor: HU-P0 85 K - trailer: HU-P1 85 K

) B Owner: - Firma Transnuklear, Hanau

C) Dimensions : - Sifa-total length: 14.38m m 5 - Sifa-total height: 4.00m 0m - Sifa-total breadth: 2.49m 0m - loading space: - length: 6.070 mm - breadth: 2.100 mm - height: 2.300 mm

) 0 Weights: - empty weight (total : )33.73 g k 0 - usable weight: 4.270 kg for an allowable total weight of 38 t n allowabla r 8.27fo g k 0e totalt weigh2 4 f o t 13.770 kg for an allowable total weight of 47 t

Further details concerning dimensions/weights/axle weights/turning circle can be taken from the attached data sheets.

"SI FA" DATA

90 880Q_

"SIFA" MAIN DIMENSIONS

"SIFA" TURNING CIRCLES

91 Allowable Allowable Allowable weights weights we i ght s (kg) (kg) (kg) Description

total weight Sifa 38000 62000 47500

empty weight (total) 33730 33730 33 730

usable weight 4270 8270 13 770

axle weight A 1 7500 7500 7 500

axle weigh2 A t 8000 8000 10000

axle weioh3 A t 8000 8000 10000

axle weigh4 A t 8000 10000 10000

axle weight A 5 8000 10000 10000

K 1 9730 9730 13730

K 2 9730 9730 13730

S 1 U 500 14 500 14500

S 2* U 500 18500 18500

empty weights A1 3 +2 A A GES

moto t rar r cto 7200 6375 13 575 K1 A4 A5 GES

trailer 5730 U 330 20050

* when crane load fully extended

"SIFA" WEIGHT SURVEY

92 Appendix N-1.2

TRANSPORTATION OF FRESH FUEL ELEMENTS JAPANESR FO E RESEARCH REACTORS

. KANDAK . NAKAGOMY , E Research Reactor Institute, Kyoto University, Osaka, Japan

Abstract Administrative procedures in Japan for transportation of fresh fuel elements are described.

1. Introduction

e transportatioTh f nucleao n r material generalls i s y assorte threo t d e means; road transport a transpor ,se r transport ai d an t n JapanI . , fresh fuel elements (enriched uranium fuels) for research reactors are ususally transporte y vehicleb d s froe fabricatioth m ne reacto planth o t tr site when the plant is located in Japan. In the case of foreign fabricators the transportation of fresh fuel e curreny airb th r n element.o I ta statuse s carriey i sb f nuclea so t ou d r materials transportatio Japann ni r transpor,ai s mori t e difficult thaa se n transport. Moreover e package s th veri f yt ,i e typ i ar difficuls B e o carrt t y out the transportation by air. Only one case of air transport of fresh fuel elements was experienced in Japan, which was a transportation of type A fissile clas I packageI s s from France (CERCA Japao )t n (Research Reactor Institute, Kyoto University).

In this paper we describe the administrative procedures in Japan for the transportation of fresh fuel elements.

2. Administrative Licensing Procedures

2.1 Regulations

Transportation of nuclear materials in Japan is regulated almost by the Scienc d Technologan e y Agenc yMinistre (STAth d Transportatiof an )o y n (MOT) regulations which are based on the Law for the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors (for road transport), the Ship's Safet w (fo La ya Transport rSe e Civith ld )Aeronautican w (foLa sr rai transport). These Japanese regulation e IAE e baseth A ar s n o Regulationd r fo s the Safe Transpor f Radioactivo t e Materials (1973 revised edition) r roaFo .d transport, the vehicle transportation is further regulated by the Police Agency Regulation if the packages exceed certain criteria.

2.2 Application and Approval

An example of actual administrative procedures for nuclear material transportation is shown in Fig. 1. It is the case for type B and/or fissile class nuclear material packages.

93 Application for Package Design

(Safety Examinatio y Technicab n l cO Advisory Committee) 1 3Tl CO DO •H Package Design Approval cd ———— « ———————i ^ UJ •H 4-1 •rl ————— M, 60 o* Application for Confirmation of Transportation ü O M M Package : Nuclear Safety BureauA ,ST Transportation Method

Road Transport Road Transport Bureau or Railway Supervision BureauT ,MO Sea Transport Ship Burea d Maritiman u e Safety Agency, MOT Air Transport Civil Aviation Bureau, MOT

(Examinatio y Competenb n t Authorities)

Confirmatio f Transportatioo n n Metho d Packagan d e

Transportation

. FigLicensin1 . g flo e transportatiow th char r fo t f typo n B and/oe r fissile class nuclear material package.

. 3 Experience

r transporai Whe e e carriew nth f freso t ou dh fuel elements (45 % enriched uranium; MEU) from France to Japan in 1981, the IAEA certificate of competent authority of the United States was required because of transient stop of the cargo plane at Fairbanks Airport in Alaska.

94 Appendix N-1.3

THE UKAEA UNIRRADIATED FUEL TRANSPORT CONTAINERS

. PANTER R United Kingdom Atomic Energy Authority, Harwell, Didcot, Oxfordshire, United Kingdom

Abstract

The UKAEA transport containers for unirradiated MTR fuel elements are briefly described.

R fueMT l elements manufacture U.Ke th .n i ared , prio o irradiat r - tion, transported by road or air using the type GB/1612A container or the shorter type GB/3104A container.

These container f similiao e ar s r construction, being rectangular lidded steel boxes, lined with core slab d usinan s g synthetically bonded hair packing to support eight elements in two layers. The type 1612 is illustrate e attacheth n i d d diagram.

The containers have been approved in 1982 and 1983 respectively unde e 197th r 3 IAEA regulations s typ,a e B(U)F fissile clas I designsI s .

95 MTR Type -fuel transit container Desig . 161nNo 2 Reference Drawing No. FE 10758

General Description No. of Flasks Non pu light mild rtMl contain«' with r»mov»bl« lid. 29. Unladen Weight

Materials (shielding) 0.128" mild itMl. Cavity size or capacity 6'-« r-Ox " V-3K"x " 8 MTR lull «lament».

FUER MT L 6 ELEMENTS

,'-7V 2-6"

96 Appendix N-1.4

TRANSPORTATIO UNIRRADIATEF NO D TRIGA-LEU FUEL

GA TECHNOLOGIES, INC. San Diego, California, United States of America

Abstract

Shipping container r unirradiatefo s d TRIGA-LEU fuee ar l described.

Unirradiated TRIGA fue s shippei l n licensei d d shipping containers designate eithes a d r TRIGA- TRIGA-2r 1o e TRIGA-Th . 2 containes i r designe speciar fo d l elements suc s fuel-followea h d control rodd an s temperature instrumented fuel rods.

Seven 1.5 in. nominal O.D. fuel elements or 25 of the 0.5 in. nominal O.D. rods fit in the TRIGA-1 container.

Descriptions of the shipping containers are as follows:

Mode TRIGA-1. No l .

Description: TRIGA fuel element shipping container e outeTh . r packaging is fabricated to DOT Specification 6J requirements. The outer dimensions are approximately 22.5 in. in diameter by 36 in. high e inneTh .r vesse a 5-in s i l . Schedul carbo0 4 e n steel pipe. Dimensions of the inner vessel are approximately 31 in. in height wit a h1/4-in . thic k5-ina wald an l. insidef o diameterp to e Th . the inner vessel is a threaded pipe cap and the bottom is a welded 1/4-in. thick flat disc. The inner vessel is centered and supported within the outer packaging by eight, 3/8-in. diameter braced, support spacer rods e voiTh .d betwee e inneth n r vessed an l

the outer packagin s fillei g d with veriniculite tampe minimua o t d m density of 4.5 lb/ft.3 Maximum gross weight including contents is approximately 235 Ib.

97 Model No: TRIGA-2.

Description: TRIGA fuel element shipping container e outeTh . r packaging is fabricated to DOT Specification 6J requirements. The outer dimensions are approximately 22.5 in. in diameter by 55 in. high. The inner vessel is a 5-in. Schedule 40 carbon steel pipe. Dimension e inneth rf o svesse e approximatelar l n heighi . tin 0 5 y with a 1/4-in. thick wall and a 5-in. inside diameter. The top of the inner vessel is a threaded pipe cap and the bottom is a welded 1/4-in. thick flat disc. The inner vessel is centered and supported within the outer packaging by eight, 3/8-in. diameter braced, support spacer rods e voiTh .d between e inneth r vessed an l the outer packagin s fillei g d with vermiculite tampe a minimu o t d m density of 4.5 lb/ft . Maximum gross weight including contents is 3 approximatel. Ib 0 33 y

98 Appendix N-2 TRANSPORTATION OF SPENT FUEL ELEMENTS

Appendix N-2.1

REMARKS ON THE TRANSPORTATION OF SPENT FUEL ELEMENTS

. KRULW L GKSS — Forschungszentrum Geesthacht GmbH, Geesthacht, Federal Republi f Germanco y

Abstract

Informatio d date providean nar a n severao d l aspecte th f o s transportation of spent fuel elements. These aspects include contract, transportation, reprocessing batch size, and economical considerations.

1. Contracts contracA ) reprocessinr a fo t spene th g t fuel elements - witUS ha Departmen Energf o t y (DOE) reprocessing plant (e.g. Savannah River, Idaho). Present contracts will be valid till 31.12.1987.

e attachmenTh thif o stA preparee b contrac o t eacr s fo dh ha t transport and sent to the reprocessing plant. After this one- receive acceptance th s actuae th f leo transport e acceptanc.Th e is valid for one year.

Remark: Standard contracts include onlU (enrichmenyHE ) % 0 2 t>

and UA1 U3Ûr o X 8 fuel other .Fo r fueothed lan r enrichments, contaco t ons eha t US-DOE directly suicidd an U .LE e fuels will probably be accepted for reprocessing in the near future.

b) Transport company

e transporTh t company coordinate e transportatioth s e th r fo n whole route (e.g homt .abroadd a ean , harbours, container ship, action physicar fo s l security, reprocessing contract). Normally the transport compan s subcontractoryha e transportath r sfo - tion.

99 c) Risk insurance

The needed risk insurance differs from country to country. This is of main importance if during shipment the container ships entry port f thiro s d countries.

. 2 Transportation

) a Minimum decay tim f speno e t fuel elements: normallr o 0 12 y 200 d.

) b Cranes

When comparing the max. laden weight of the casks with the needed allowable somd cranad eo et weight weights ha se - ,on wa (e.g . ter insid e casksth e , ropes, impact absorbers, fuel elements). ) c Subpressure

The casks have to be dried by air. After drying inside the container, the pressure should be ca. 6 x 10^ Pa.

d) Contamination

The cask surface has to be free of contamination 10~£ Y 5ß+ uCi/cm2 a10~£ 6 uCi/cm2

e) Cask limitations

There exists cask limitations for U-5 content of the fresh fuel elements, decay heat, fission product inventory.

) f Physical protection

n manI y countrie physicae th s l protection demand r transporfo s - tatio f spenno t fuel element e highesar r thae IAEth nA recommen- dations mann I . y cases guarded transportatio d additionanan l communication systems are required.

100 g) Shipping

e pric Th r shipmenfo e t with container ships depends mainln yo volume e othee th th n ro n th.O t e no caske weigh th d san f o t hand,US-regulations require that splittable cargos be transported independently. Therefore, it is recommended to ship two casks with two containers. Otherwise, one has additional cost in the US-harbour for separating the casks onto two containers.

. 3 Reprocessing

a) Minimum reprocessing batch

One reprocessing batch consists normally of that number of fuel elements which have been transported to the reprocessing plant within 60 days. On the other hand, the reprocessing price has two parts:

100. ca 0 - US-$/kg (U+A1 198n )i 5 - a minimum charge of 44.500 US-$.

Therefore a reprocessing batch should consist of sufficient spent fuel elements thae totath t l weigh greates i t r - thami na nimum weight (contact DOE or transport company). Otherwise, it is recommende storo t de fue th el element e reprocessinth n i s g plant till the total weight is higher than this minimum weight. It is possible that smaller reactor stations can reprocess their spent fuel together. But in this case, the reprocessed fuel can only be separated by calculations. Additional fees need to be pair storagfo d f speno e t fuel, conversio UFgo nt , shipmeno t the enrichment plant.

) b Spent fuel Taking 1000 MWd/a, 40 % burnup, 200 g U-235/fuel element, a ratio of 1 : 5 for the number of control fuel elements and fuel elements, there wil producee lb d aroun spen7 1 d t fuel elements per year at the facility.

101 . 4 Economical considerations

a) Transportation Transportation cost depen : managemenon d t cost, insurance, physical protection demands, amortisation for the casks and shipment. The shipment cost depend mainly on the freight rate of the container ships (see 2.e) .cranef I s allow transportation with heavier casks wilt , i cheape e le b thes us o et r caskse th s ,a possible number of fuel elements which can be transported increase faster than the cask weight.

) b Reprocessing Reprocessin reasonabls i g reacn minimuca e th he on onl mf i y weight for a reprocessing batch (see 3.). If there are enough spent fuel elements at the reactor facility, the reactor operator has to check the possibility of cutting off the upper and lower parts fuee (Alth l f )elemento s sinc reprocessine th e g price depends on kg (U + Al). Then there have to be at the facility cutting tools, special handling tool d storagan s e possibi- lities for the Al-waste.

102 Appendix N-2.2

TRANSNUKLEAR SPENT FUEL SHIPPING CONTAINERS

TRANSNUKLEAR GmbH Hanau, Federal Republic of Germany

Abstract

Detailed date provide TN-7e ar a th n ,o d TN-7/2, Goslard ,an TN-1 spent fuel shipping container r MTR-typfo s e fuel elements,

103 ——U — j* 1360 ——— -j , ^ [IN/ -~ö '"T" ' ' — UHe*a .J - N M •* : r / ;:;

^c'-*^_ ' I >m^ - sue k l«t«« l 0 2 ) X • I ; 1 1 t stee^ ^ «tee t m • < 2 ' « ' * ti ( ! i B : i " ' "inewletic^aT^« T-" , ? « ..'•'. s« <' CQ H -, ' insulatio< b P n'• 130 i, 30 : ' : , C l; 3H ^::ik^ . , *— h -iEfcii ; , j . a T ! . ,l70PI b additional

TN-7 TN-7/2

max. laden weight (tonne») 25 22 weight during handling (tonnes) 24 20

interior dimensions (mm) at 540x2650 ••54 2680x 0

authorisation B(U) B(U)

package approva . (PTBno l ) D/4001/B(U)F (-)

nuclear fictile class II I!

category - yelloID w 01 -yellow

f containero . no s available 1 sinc1 e mid. 1982 f Inneo . rno basket r containepe s r 4 4

method of loading/unloading under water under water

MTR fuel elements: O D 1 0 MERLIN plpp MERLIN

- capacit container pe y r t cu 4 6 t cu 0 6 t cu 4 6 t cu 0 6

- element dimensions (mm) 0-95x632 80x86x658 ,»95x632 80x86x656

- orlg.U-235cont.per elemen) I' t ( g)0 27 ) 18(- 5 185 -) 270 (->

- max. orlg. enrichmen) % ( t 3 9 (20 3 (209 )) 3 99 3 (20) )

- max. decay heat per container (kW) 4.5 4.5 4.5 4.5

M TR container details

104 Goslar TN1

Goslar TN-1

HMX. taden weight (tonn**) 11 1« during handling (tonn««) 10.2 17.8

Interior dimension* (mm) /ir 403 i 960 240x 0 Omi42 n*

author l «et Ion B(M) B(U)

package approval no. (PTB) D/4O53/B(M)F D/4OO4/B(U>F

nuclear flaaile claa« II II

category III-yellow II - ye How or HI - yellow

f containero . no « available 2 1 f inneo . rno backet r containepe « r 1 3

metho f loading/unloadindo g under water under water

MTR fuel elements:

- capacit container pe y r 13 42 cut

- element dimension« (mm) 80x86x950 79 «86x670

- orlg.U-235cont.per elemen g( t ) 320 - 400 (-)

- max. orlg. enrichment (%) 93 (20) 93 (20)

- max. decay heat per container (kW) 3.2 5.4

MTR container details

105 Appendix N-2.3

UKAEA'S 'UNIFETCH' IRRADIATED FUEL TRANSPORT CONTAINERS

R. PANTER United Kingdom Atomic Energy Authority, Harwell, Didcot, Oxfordshire, United Kingdom

Abstract

The UKAEA "UNIFETCH" containers for transport of irradiated fuel from research reactor e describedar s .

In the UK, irradiated fuel elements from research reactors are transported by road using the UNIFETCH containers.

These container e finnear s d cylindrical steel containers f whic,o h versionso tw ther e UNIFETCe ar e th , H 'H1, typ. GB/1112ANo e , suitable for croppe R typMT de elementlongee th d r an sUNIFETC H 'L', typ. No e GB/1113. The type 'H' has a 26 element basket, while for the type 'L' there are alternative 24 and 40 element baskets.

The containers are loaded and unloaded under water, but the element e transporte ar sy state dr a .n i d

These containers were originally approved under the 1967 IAEA regulations as a Large Source Package design, but have in 1985 been reapprove s typa d e B(M)F fissile Clas I containersI s .

e attacheTh d illustration d datan sa sheets give sizeth e , weight and handling details.

107 'Unifetch' Typ Transpor' H e% t Container Design No. 1112 Reference Drawin176H D . 7 gNo f Flasko . sNo General Description Gas tight (test prtuure 150 pj.i.g. hydraulic) vertical 1. cylindrical finned M.S. container with removable lid. Unladen Weight

Primarily intended for the transport of irradiated M.T.R. 15 to)M 3 ewts. 3 qtrt. (without inner basket. Design No. 1423). type fuel elements, using inner container Design No. Materials (shielding) 1423- 12%" mild steel. Cavity size or capacity 2'-6" dia 2'-5K.x "

Safe Loadin f Liftingo g Points 22.5 tons.

Approved Lifting Harness Drawin. gNo 1767/00H E 5 (Lifting 1767/00 H frameE d an ) 1 (Lifting ear). Both lift flask complete with hold-down equipment. Total weight 18.52 tons.

Max. Loading of Harness Lifting fram tons5 2 e— . tons5 2 — Liftin. r gea

Lifting Harness Plant Item No.

Vehicle Any suitable and approved vehicle also transported by rail and sea.

Approved Hold Down Equipment Drawing No. By road - rail and'sea - EH 1767/003.

Speed Restrictions 5 M.P.H. on site

Normal Storage Flask, storage compound

Routes Primarily intende internationae th r dfo l traffi f M.T.Rco . type fuel elements.

Normal Usage Transpor f M.T.Ro t . fuel elements from varioud an . sK siteU n si abroa do O.N.P.O.Et .

Ancillary Equipment Inner container, desig . 1432nNo ,767/ 1 drawinH E 8 . gNo

108 Design No. 1112

2-6* DIAM

s-10 Va* -JO s-a

109 'Unjfetch' Type 'L' Transport Container Design No. 1113 Reference Drawin 23156E A . gNo 3

f Flasko . sNo General Description M tighG t (test pressur 0 p.s.i.g15 e . hydraulic). 2 vertical cylindrical finned M.S. container with removable lid. Unladen Weight Primarily e Intendetransporth r f fo do irradiate t d 16.8 tons (without inner basket). M.T.R. type fuel elements, using inner container Materials (shielding) Design No,. 1331. 1376, and 1753. ^.. m||d ,„„ Cavity size or capacity 2'—6" dia. x 3'—5" long (approx.).

Safe Loading of Lifting Points Ancillary Equipment 22.5 tons. Inner container, desig 133. nNo 1 - Drawin ZA. gENo 60705. Inner container, design No. 1376 — Drawing No. AE 231573. Approved Lifting Harness Drawin. gNo «>"«'"<". <*•**"° - Drawin g NO. ZAE 61218 lnner 17SN 3 (a) AE 231580 (this harness lifts flask complete with hold-down equipment when being trans-shipped). Total weight 21.1 tons (approx.). (b) AE 231585 (lid removal).

Max. Loadin f Harnesgo s tons5 2 (a) . ton(b8 ) s (lid removal).

Lifting Harness Plant Item No.

Vehicle suitably f sitAn Of e• approved ean d vehicle. — also transported by rail and sea. 25 ton 'Carrimore' trailer (on site only).

Approved Hold Down Equipment Drawin. gNo By road - rail and sea - EH 1767/003. 25 ton 'Carrimore' trailer (on site only) ZAE 61075.

Speed Restrictions 5 M.P.H. on site.

Normal Storage D.E.R.E. flask storage area.

Routes Primarily intended for the international traffic of M.T.R. type fuel elements.

Normal Usage 1. Transport of M.T.R. fuel elements between various sites in U.K abroad an . D.N.o dt P D.E. . 2 Transpor f F.Rto . breeder slugs fro m120D Windscaleo 6t .

110 Design No. 1113

5-10

111 Appendix N-2.4

THE TRANSPORT OF SPENT FUEL ELEMENTS OF RESEARCH REACTORS

COMMISSARIAT A L'ENERGIE ATOMIQUE Institu recherche d t e technologique développemene ed t t industriel, Division d'exploitatio s réacteurnde s prototypes t expérimentauxe , Servic piles eSaclayde e sd , Centre d'études nucléaire Saclaye sd , Gif-sur-Yvette, France

Abstract

' caske 04 COGEM Th u , 'l Afrequentl y calle e 'Pegaseth d ' s availablcask it d som an f ,o e e internal containere ar s briefly described.

Spent fuel elements of research reactors are transported between the different French research centers, or between one of these centers and e reprocessinth g plants handling these fuel n Belgiumi s n Franci , d an e finally in the USA, in a single model of transfer cask, of which many copies have been built. Thi U 04's'I cas,e knows th i kfrequentl s a n y >* called 'Pegase' cask ,e firs froth e nam th tf m o e reacto r whicfo r t i h was commissioned.

Onle cask'th y s internal arrangement r baskeo s t varies froe reactoon m r to another.

The 'IU 04' casks are owned by COGEMA, which maintains them and makes them available to users.

Container AA4r Osirifo 9 s fuel elements, consistin f fiv° o gcenter 72 e - angle segments, capable of accommodating thirty elements.

• • * Container AA5 r Siloe0fo , Triton, Melusine fuel elements etc., ° center-anglconsistin60 x si f o g e segments, capabl f accommodatino e 6 3 g elements.

113 Containe 3 reactoEL e rth rAA7 r cells fo 0 7cavitie 12 : r containerpe s .

Containe F elemenGrenoble RH th n f a ro t r eAA11 fo ILL7 .

This list is not exhaustive.

assemble e weighth Th f o t y equippe r transporfo d d loadean t d with fuel elements varies slightly with loading e orderTh . magnitudr o s s a e ar e follows :

• cask: 17,500 kg, • transport equipment (cove frame)d an r : , 310kg 0 • internal container and fuel element load: 3000 kg (2500 to 3500 kg),

This make a totas l weigh f abouo t t d possibl23,60an , 0s kg hig a ys a h 24,00. 0kg

114 WEIGHTS MAIN BODY COVED LI R FRAM£ TURNBUCKLE

3 PACKAGING LIFTING LUG

BODY LID TOTAL . 17500 kg

COVER LIFTING LUG

4 TRIMMING LUG

2 LIFTING GLU

STAINLbSS STfcBL

IU-O4

115 /V/\-49 "PEGASE" FUEL ELEMENT CASK 0TO rfbIGHTS g K E BASKEO ON 4S : T ChNTRAg K L0 BOD10 : Y TOTAL : ZSOO Kg

ONE BASKET LIFT SCREW

S BASKET Y IU-0B S 4

EACH E FUEBASKEON LR FO T CONTROL ELEMENT S BASKlsTi 84x84 mm

PAINTING COOPER

AA-50 "SILOE" FUEL ELEMENT CASK 070 WEIGHTS ONE BASKET : 316 kg CENTRALg K BODY 0 10 : TOTAL : 1996 Kg

ONE BASKET LIFT SCREW 6 BASKET IU-0Y B S 4

EACH BASKt86,5x77,6 R FO T 5 LOCATIONS 150m m 7

PAINTING COOPbR ————-

116 Appendix N-2.5

TRANSPORTATIO IRRADIATEF NO D TRIGA-LEU FUEL

GA TECHNOLOGIES, INC., San Diego, California, United State f Americso a

Abstract

Containers for shipping irradiated TRIGA-LEU fuel in the United n EuropStatei e describedd ar ean s .

Shipment e Uniteth n di s State f irradiateo s . nominain 5 1. ld O.D. TR1GA fuel have been made in the BMI-1 shipping cask currently owned by Cintichem, Inc. The fuel shipped in the cask has come from reactors which e containeoperateth d an t powe a drW M carrier2 levelo t p u dp u s 8 elementst3 o A descriptio. s followsa e cas s th i k f o n.

Mode ; BMI-1No l .

Description: Steel-encased lead shielded shipping cask. The basic cask body is a cylinder 33.37 in. in diameter by 73.37 in. high o concentriformetw y b d c stainless steel shells whose annular region is filled with lead. The outer 1/2-in. thick shell has a 0.12-in. thick plate spot providin, weldeit o t d 0.06-ina g . thick air gap insulator. The inner shell is 15.5 in. inside diameter by . insidin 4 5 e stainlesa lengths i e casd Th li .ks steel weldment having f leao 7.7 d. 5in shielding s securee i e casth d Th o .li kt d cask by twelve steel studs which are welded to the cask body. Cask appurtenances includ a draie n line with needle valv d plugan e , pressure gaugepressura d ,an e relief valve e totaTh . l cask weight, including maximum contents of 1,800 Ibs, is 23,660 Ibs.

Shipment Europn i s f irradiateo e . nominain 5 1. ld O.D. TRIGA fuel have been mad y Transnukleab e e Goslath d TN6/ n an ri r 3 type contain- e GoslaersTh . r containe ra capacit weighs 0 fue6 ha 0 tonf ld 1 so y an s elements. The TN6/3 container weighs 6.3 tons and has six positions for special elements suc s fuel-followea h d control rodd temperaturan s e instrumented fuel elements. The fuel shipped in these containers has come from reactors which operate t powea . d MW r1 levelo t p u s

117 Appendi3 xN- SPENT FUEL STORAGE

Appendix N-3.1

NUCLEAR CRTTICALITY ASSESSMENT OF FUEU HE L LED ELEMENUAN T STORAGE

R.B. POND, J.E. MATOS Argonne National Laboratory, Argonne, Illinois, United States of America

Abstract

Criticality aspect storinf o s (93ZU (20%U HE gLE d ))an fuel elements have been evaluate functioa s a d 235f no U loading, element geometry fued ,an l type. Suicide, oxide alumid ,an - nide fuel types have been evaluated ranging in ^35rj loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed for- mations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data.

INTRODUCTION

Criticality aspects of storing fuel elements is of concern to all reactor operators. Any change to the types of fuel elements approved for storage may require that the subcriticality of a storage rack be reconfirmed. As an insight into what migh expectede tb , this report presents result studa f so y assessing the storage of HEU and LEU fuel elements with various fissile contents.

SCOPE

Fuel Element Storage Model

assessinn I g fuel element storage type th ,f fuee o l storagelemene th d etan configuration must be defined. For purposes of this report, twenty fuel element types were considered and a generic storage rack was used.

The storage rack is defined as an unpoisoned aluminum framework within which partitions form individual fuel element storage compartments. The entire unit is immersed in a pool so that the fuel elements are moderated and reflected with water. An infinite-by-infinite array of fuel elements is assumed and the separation between storage compartments is adjusted to control the storage rack reactivity. On this basis, reactivity effects associated with various types of fuel element varioun si s storage rack configuration madee b n . sca (The bases for this fuel element storage model are developed in Appendix A.)

119 Calculation Model Three-dimensional (XYZ) diffusion theor calculatione th s use i yn i d n i s this report. In many of the calculations a simplified representation of the storag e fueeth l rac d elementan k usede sar . These simplifications included neglect of the aluminum storage compartments and the fuel element end-fittings. Sensitivity studies to assess the reactivity effects of these simplifications were, however, made and validation of the diffusion theory calculations were made using Monte Carlo techniques.

CROSS SECTIONS

Microscopic cross sections for the fuel elements and the storage rack were calculated usin e EPRI-CELth g L code* with ENDF/B-IV cross section datae Th . Integral transport calculations in EPRI-CELL are performed for 69-fast groups and 35-thermal groups «1.855 eV), and then collapsed to 5-broad groups with upper energy boundarie MeV0 1 ,f o s0.82 1 MeV, 5.53 e keVTh 0.62,d . an 1.85 5eV , 5eV fuel element geometre unit-celth d an y l model EPRI-CELe s th use n i d L calculations are show n Figi n. 1 .

Fuel Eieaent Specification»

(S«e detail!r Tablfo 1 e ) Broad group cross sections Vl/Ulfl/l were generated for each fuel ele-

TOo outermostv e t plates hava e ment type using the flux spectrum clad thlcknea f 0.049o c . em 3 of the core portion of the fuel element. The core portion of a

l dlnenslonAl cm.n i s fuel element include e fuelth d , clad watean d r channel regions a s show Fign i n. 1 .

Separate microscopic cross sectione fueth l r elemenfo s t sideplates were generated using a pure 235u fission spectrum on 80/2a 0 volume percent mixture of aluminum and water. The side-

Fuel Element Unit-Cell Modela plate portion of a fuel element included the portion of clad on Reflective boundary conditiona on ttie coie unit-cell facea. the fuel plates between the fuel e l*,0 MOOtRATOR mead sideplatean t s plue samth se H - 0.0 II •— — —2 corresponding part of water in

Core Unit-Cell. the water channels. Macroscopic cross sections appropriate for the sideplates of each fuel ele- Volume Fraction toit Cell A1/H20 ment type wer e neue th use- n i d Hack 100.0/0.0 tronic calculations. ———— nasion Sidea «0.0/20.0 Spectrum Enda SO. 0/50.0 The same methodology as Reflector 0.0/100.0 used for the sideplates was used in generating cross sections for the fuel element end-fittings, the aluminum rack and the storage Fig . 1 .Fue l Element Specificationd an s rack water reflector. Unit-Cell Models.

120 FUEL ELEMENT DESCRIPTION

Thirtee (20U nLE Z enriched seved )(93U an nHE Z enriched) fuel element type usee thin sar i d s study coverin wida g e rang fuef eo l densities, fuel types fued ,an l element geometries choice Th .f fue eo l element type made sar e based upon types plate-typn currentli e us n yi e researc tesd han t reactorsd ,an types which migh expectee tb available b o t d s fueea l material technology develops e fueTh .l element geometries considered contain3 2 betweed an 6 n1 plate elementr spe . Detailed specification twente th f yso fuel element type listee ar s n i d e rangTablTh fuef eo . el1 densities, fuel types fued ,an l element geometries are summarized below.

Plates per Fuel Meat Uranium 235U Fuel Type Element Thickness, nun Density, f/cc Loadingg ,

LEU Ü3S1-A1 19 0.51 3.1-5.3 225-390 23 0.51 3.2-7.0 280-621

LEU U30g-Al 23 0.51 3.1 278 16-22 0.76 3.1 288-396 20 1.00 3.1 473

HEU DA1X-A1 19 0.51 0.5-1.2 180-405 23 0.51 0.4-1.3 180-530

Table 1. Fuel Element Loadings Fuel Sldepltt« Eleaent Uraniuta Water Voluae FractionsZ , Loading Fuel Platea/ »=u/ Demlty. Fuel Hea) (F t Channel (M)

Ktinber Type Element Elenenc, g g/cc Thicknessa n , Tblcknea«, ma Al H20

1 LE3 U2 UjOg-AJL 278 3.130 0.51 2.188 81.11 18.89

2 LEU6 1 U308-A1 288 3.130 0.76 «5. 3 1 79.56 20.««

3 LEU U308-A1 18 32* 3.130 0.76 2.899 80.56 19.««

4 LE0 U2 Uj08-Al 360 3.130 0.76 2.457 81.51 18 .«9 5 LE2 U2 UjOg-Al 396 3.130 O.76 3 .« 2 « 2.09S 17.57

6 LE0 U2 OjOj-Al «73 3.130 1.00 2.217 83.0« 16.96

7 LEU UjSl-Al 19 225 3.071 0.51 2.916 79.49 20. SI 8 LEU UjSi-Al 19 350 «.778 0.51 2.916 79.49 20.51 9 LEU UjSl-Al 19 390 5.32* 0.51 2.916 79.49 20.51

10 LEU3 2 U3S1-A1 280 3.157 O.S1 2.188 «1.1118.89

11 LE3 U2 U3Si-/U 320 3.609 O.S1 2.188 81.1118.89 12 LEU UjSl-Al 23 390 «.398 O.S1 2.188 81.11 18.89 13 LEU UjSl-Al 23 «21 7.000 O.S1 2.188 81.11 18.89

1« HEU UA1X-A1 19 180 0.528 O.S1 9 .« 9 7 2.916 20. SI IS HEU UAlg-Al 19 280 0.822 O.S1 2.916 79 9.« 20 .SI 16 HEU UA1„-A1 19 «OS 1.189 O.S1 9 .« 9 7 2.916 20 .SI

17 HEU UA1X-AI 23 180 0.437 0.51 2.188 81.11 18.89 18 HEU UAlj-Al 23 280 0.679 O.S1 2.188 81.11 18.89

19 HEU3 2 UA1X-A1 «OS 0.982 0.51 2.188 81.11 18.89 20 HEU UAlg-Al 23 S30 1.28S 0.51 2.188 81.11 18.89

121 All fuel elements are assumed to be fresh in accordance with standard practice for this type of criticality assessment. The presence of any burnable poison which might be required in many of the heavier loaded fuel element alss si o neglected. These assumptions about fuel element poisoning effects are made in order that all calculated reactivities for the storage rack configurations will be conservative. The fuel elements are assumed to lonm c activm gc 8 b 6 ewit0 6 e h a fue abovm l c belod heigh4 an e d wth an t fue simulato lt e fuel element end—fittings.

STORAGE RACK CALCULATIONS

The first part of this section examines the reactivity trends that one fuel element type will hav varioun ei s storage rack configurationse th n .I second part reactivite th , y trends thastorage ton e configuration will have with each of the twenty fuel element types are examined.

Overall, the results of this section are intended to provide the means of estimating eigenvalues for various fuel element types in various storage rack configurations. Based upon these data, a reactor operator will have a basis upon which to estimate the reactivity effect of substituting one fuel type for another in unpoisoned storage racks. (Examples illustrating the use of these datprovidee ar a Appendin i d ) xB.

e storagTh e rack mode calculatione l th use n i d s assume n infinite-bya d - in-finite array of fuel elements in which there are an infinite number of fuel elements in a row and an infinite number of rows. The spacing between fuel separatioe th d an elementw nro betweea n si n row specifiee sar defino t d e th e storage rack configuration. The calculational models used in this study are show Appendin i n . xC

Eigenvalues for a Given Fuel Element Type in Various Storage Rack Configurations

Eigenvalue results Tabl . e2 Eigenvalu e Calculation Infinite-byn a r sfo - are shown in Table 2 Infinite Arra Numbef yo Fue3 1 r l Elements for an array of LEU (LEU 235 g FunctioU a U 1 Ss )i62 a f no 235 3 U3Si 621 g U fuel Element Separation in a Row and Row Separa- elements with row sep- tion. (See Fig Appendin i . D ) xC. arations of 10 to 22 cm, and element separations in a row of 0.0 to 2.0 cm. Elèvent These data are plotted Sep«r*tlon Row Separationr ctt in Fig. 2 as a function • Row e l , CM 10 12 14 16 18 20 22 of element separation 0.0 .0419 0.9704 0.9206 0.8868 0.8643 0.8494 0.8398 for various row separa- 0.25 .0398 0.9691 0.9198 0.8865 0.8642 0.8496 0.8401 tions. In these cal- 0.50 .0360 0.9662 0.9176 0.8847 0.8628 0.8483 0.8390 culations, both the 0.75 .0309 0.9620 0.9141 0.8C16 0.8600 0.8456 0.8365 aluminum of the stor- 1.00 .0245 0.9565 0.9092 0.8772 0.8559 0.8419 0.8327 agfuee eth l racd an k 1.50 .0078 0.9417 0.8958 0.8647 0.8441 0.8305 0.8216 element end-fittings 2.00 (1.9868 0.9226 0.8782 0.8481 0.8281 0.8149 0.6064 were neglected.

122 The result n Figi s .2 show the relative reactivity effects for this fuel • ^(laUn T KPMAtiai

DATA FD* »nWT€-»Y-«fWII£ element functioa typ s a e n AHMY FIMCTIOA S XA tOF NO V JtPAMTKHI. of various storage rack con- FUEL PLATO PMULLtL ID «0»S. figurations. Interpolations to determine the eigenvalue for any specific configuration can also be readily made. The Oak Ridge Research (ORR) reactor storage rack2, for example, has a row separation of 17.2 cm and an element separatio f 1.7no . 7cm According to Fig. 2 the eigenvalue for this configuration wit e LEth hU t^Si 621 g 235U fuel elements would be 0.8431. This (calculated) eigenvalus i e plottes i Figd n I dan .2 identified "ORR". 1.00 ISO ELIKUI XPARATKM a Fig. 2. Eigenvalues for Various Row Separations in Infinite-by-Infinite Arrays of LEU UßSi 621 g 235U Fuel Elements as a Function of the Separation Between Elements.

Eigenvalue a Give r fo sn Storage Rack Configuration with Various Fuel Element Types

Tabl . 3 eEigenvalu e Calculation r Twentfo s y Fuel Elements Loadings, Eacn i h an Infinite-by-Infinite Arra Assumind an y e ORth gR Fuel Storage Rack Spacing Specifications of 1.766 cm Element Separation and Separationw Ro 17.2m c 4 . (See Fig n Appendii .E ) C. x fuel El taint Loading ?u«l Pl.t««/ »S0/ k«ff Hua her typ« Element Elenent, g Eigenvalue

1 LEU U3<>8 23 278 0.7410 Eigenvalue resulte ar s 2 LEU U308 16 288 0.7587 shown in Table 3 for infi- 3 LED 0303 ia 324 0.7695 nite e arraytwentth f o ys 4 LEU »303 20 360 0.7765 fuel element typea n i s 5 LEU U 0 22 396 0.7803 3 8 storage configuration having 0.7967 6 LED U30g 20 473 e fueth l element spacing 7 LEU 0351 19 225 0.7154 S LEO »351 19 330 0.7889 specificationR OR e th f o s

9 LED U3S1 19 390 0.8044 storage rack. These data are plotted in Fig. 3 as a 10 LEU U Si 23 280 0.7402 3 function of the 235U load- 11 LED U3S1 23 320 0.7613

12 LED U3S1 2Î 390 0.7899 inn eaci g h fuel element 13 LEU U3Si 23 621 0.8453 type n thesI . e calcula-

14 BED UA1X 19 180 0.6903 tions, 1/8 in. (0.32 cm)-

IS HEU OA1X 19 280 0.7772 thlck aluminum storage 16 HEU DA1Z 19 405 0.8377 compartments were included

17 , BEDDM 23 180 0.6783 and the fuel element end-

18 HEU OA1X 23 280 0.7635 fittings were neglected.

19 BED UA1X 23 405 0.8229

20 HEU UA1X 23 530 0.8594

123 The results in Fig. 3 0.90 ••" iI I 111 I I IT 11 11 I 1 I 11 M l 1 1I 1111 1 11U 1 1 1lI 11 l 11111 l l l 11 l M 11 l l 11 l sho 11relative 1 wth l l l l l Le reactivity effect f fueso l FUEL STORAGE RACK element storag thin ei s WITH 1.766 em /ELEMENT configuratio funca s na - SEPARATION AND 17.24cm/ tion of: (1) the number ROW SEPARATION of plates per element, 0.85 fuee th l ) mea(2 t thickness in an element, and (3) LEU vs. HEU fuel types. Interpolations to determine eigenvalues for other 235U fuel ele- 0.80 ment loadinge b n sca readily made.

The data in Fig. 3 indicate that reactivity fuee th effect lo t e sdu element geometry can be 0.75 characterized as a func- H/e 235th tio f Uno atom fuee ratith l f elemento . For these fuel element MEA0.5m m 1 T U U PLATE9 3S1 LE i D S geometries eigene ,th - O LEU U Si 23 PLATES values are inversely pro- 0.70 3 portional to the number • LEU U30, 0.51mm MEAT (13 PLATES) of plate elemenr spe d tan • LEU U308 0.76mm MEAT (16,18,20,22 PLATES) inversely proportional to the fuel meat thickness. UU 30LE j 1.0mv m MEA PLATES0 (2 T ) As would be expected, 0.65 I 11 I I I I I I 11 I I I I I I I 111 I II II 11 l 11 l l l 11 M I I I I I 11 11 I I 11 I I LE11U fuel is less reactive 100 200 300 400 500 600 700 than HEU fuel for a given 235 U MASS/ELEMENT, G fuel element geometrd yan 235U loading. It is also evident that the eigenvalue resultt no e sar Fig. 3 . EigenvalueU HE Variour d sfo an U sLE sensitive to the form of Element Infinite-by-Infinitn si e LEU fuel since the LEU Arrays With Separations of 1.766 cm UßSi and LEU UßOg results Between Element 17.2d san Betweem 4c n for 23-plate elements Functioa Row 235 e s sa Uth f Fueno l with 0.51 mm fuel meat Element Loading. are almost identical.

SENSITIVITY STUDIES Configuration Model

e sensitivitTh eigenvaluf yo e calculation effecto storagse t th f so e rack compartments, fuel element end-fittings, fuel element sideplatese th d an , diffusion theory model mesh have been evaluated. These results are listed in Table 4 for four storage rack configurations.

124 Table 4. Eigenvalue Sensitivity of an Infinite-by-Infinite Array of Number 13 Fuel Elements (LEU 235 g U-SFunctioa U 1 s )ia 62 Fuef no l Element and Storage Rack Representation Calculationae th d ,an l Model.

Eigenvalue Assuning the ORR Fuel Storage Rack Ceonetry Eigenvalue Aaauntng 17.24 en/Row17.64 en/Row 14 en/Row14 en/Row 1.76 / en 6 1.36/ en 6 0.2/ en 5/ 0.7 en 5 ConfiKuratlon Elenent Elenent Elenent Elenent Conclusion

J d «n , showI o 0.845 , A Tig*.n H I n , . 3C 1 0.8366 0.9185 n AppendiI C x

2. With Al replaced by H20 In the 1/8 in.-thlck Al frame region 0.8463

. 3 Withoul franA e c th tregion ' 0.8431 0.8332 0.9198 0.9141 Effec f aeso t h IS SMll

4. With an asauned 4 en-long end-fitting (45.53/55. 47-Al/H20) 0.8450 0.9182 0.9181 Effect of enda is very stall

5* Fuel only - no «idea, no frone.

no l replaceendal - a Hy 0b d 0.8SS8 0.9536 0.9391 Effec f sideo t s Is significant 2 6. Fuel and enda only - no aides. no franc - all replaced by HjO - 0.9392 Effect of ends la very saall

. 7 franFued an Ideala c o onln , y- - n replaceo1 end«1 H2- sy b d0 0.9436 Effec f frano t c is avail

olnts In the water reflector. Without the Al franc region e Y-neath , h water reflector boundaries were, respectively: 12.62(9), 12.62(9), 11.0(7) 11.0(7)d an , .

The following table summarize reactivite sth y effects liste Tabln i d e4 for one of these storage configurations; in general, all configurations show the same reactivity trends. The data are for the ORR storage rack configuration with LEU 038! 621 g 235u fuei elements.

Change in Configuration Reactivity Effect, Z 6k/k

Include fuel element end-fittings -0.03 Include storage rack compartments -0.10 Nominal (~10%) increase in mesh points* +0.32

These reactivity effects indicate that the diffusion theory eigenvalue uncertaintie storage th f so e rack configuration model lese sar s tha 6k/k% 1 n .

Infinite Array Versus Finite Arra Fuef yo l Elements

exampl n conservatise a th s A f eo m implie assuminy b d infinite-byn ga - infinite arra f fueyo l element storaga n si e configuratio finita . nvs e array, th 235 eg UßS U Ueigenvalu1 LE i62 fuee storagth R l OR r element e fo eth n si rac calculateds kwa storagR plaA .OR ne vieth e racf wo shows ki Fign ni . .4

Mesh sensitivitie svalidatioe showth n ni n studies section Indicates that the maximum mesh reactivity effect is about 0.9% 6fc/k for a 100% increase in the number of mesh points. Further increases in the number of mesh shopointt no substantia wa o sd l additional reactivity increase.

125 e racthres Th kha e rows of storage compartments and ten compartments per row. When fuel elements are cen- comparte tereth n i d - ments with the fuel element plates parallel rowse t th ospace ,th - ing between elements d an 1.7s m i c 7w ro ia n the separation between row 17.s si . 2cm The calculated eigenvalu thir efo s configuratio 0.7985s ni . This eigenvalue compares wit kha ef0.845f fo 3 (see Table 4) for an infinite-by-infinite array of fuel elements. e reactivitTh y difference is about 5% 5k/k. «U. «HUSO« M c. Calculations performed using infinite arrays Fig. 4. The ORR Fuel Storage Rack Configuration. are, therefore, clearly conservative.

VALIDATION STUDIES

Becausimportance th f o e d ofte an enecessite th n f relyino y g upon calculation determino t s e safe fuel element storage configurations diffu,e th som f -o e sion theory eigenvalues were compared with results using Monte Carlo techniques. These dat showe shows ar aA Tabln ni Tabln ni . e5 , calculation e5 s were per- formecriticaa r fo d l configuration, infinitn a f fue o r lw fo 3 ro eelements d ,an foinfinite-by-infinitn a r e arra fuef yo l elements diffusioe Th . n theory code useMontDIF3De s th wa d ed 1 *,Carlan o codes were VIMKENOd an 5 6.

Tabl . Validatioe5 f Calculationano l Methods.

Diffusion Theory ______»once Carlo Configuration DIF3D VIM KENO ENDF/B-IV EKDP/B-IV ENDF/&-IV Hanaen Roach

235 1. Critical SPERT-D HEU OA1X 306 g 0 0.9999 1.021710.0039« 0.999710.0048

23S 2. On* Infini« row LEU U3S1 621 g U, 0.25 ev/eleaent, 8 c« reflector 0.8036° 0.824410.0049 0.8353*0.0052 0.832710.0052 1.77 ça/élèvent, 6.62 cm reflector 0.7860 0.811310.0049 0.8144*0.0030

3. Inflnlte-by-lnflnlte LEU UjSi 6J1 g 23S0, 1.77 »/element, 17.2 ea/rov 0.8431 0.8642*0.0066 0.861710.0038 0.889910.0031 e

•Data are for 54K ttletoriea. Five batch«« of 30K hlitorlci each wer«: 1.0255I0.0055, 1.024910.0052, 1.022910.0046. 1.009610.0056 and 1.027410.0056.

'»Doubling the xy Matt gtvea a k.ff of 0.8092. Keah effect la of the order of 0.6X «k/k. «Doubling the zy «web give* a k.« of 0.8521. Keah effect ia of the order of 0.9Z «k/lu

126 In general, the eigenvalues calculated with Monte Carlo are systema- tically larger than with diffusion theory. The uncertainties quoted for the Monte mos n Carli td casesoan resulta l t leas± ,a o standare tw tsar d devia- tions would be required to cover the diffusion theory results. fik/% 3 o kt differenc2 e Th e between diffusion theor Montd an y e Carls oi somewhat accounted for by a mesh reactivity effect in diffusion theory. For the two subcritical configurations, this reactivity effect is worth between 0.9o t %6 0. ok/k. Based upon these comparisons, diffusion theory eigenvalues for this fuel element storage study could be underestimated by 1 to 2* 6k/k.

RESULTS AND CONCLUSIONS

e resultTh f thiso s study have shown that replacemen U fueHE lf to ele - ments with LEU fuel elements will not have a significant reactivity effect in most storage racks. The magnitude of any reactivity effect will depend upon the change in 235u loading and differences in the fuel element geometry. As an aid to assess fuel element storage, curves are developed (Figs. 2 reactivitr fo ) 3 d yan fueU effectHE l d elemenfunctions an sa U LE t f typeso s for various unpoisoned storage rack configurations. The curves cover LEU and HEU fuel element loadings between 235elemen g r approximatel U0 pe 60 td an 0 20 y with various fuel element geometries storagd ,an e rack configurations with various row and fuel element separations.

Relative to HEU fuel elements, reactivity increases associated with larger 23Su loading U fueLE ln si element scompensatee b ten o t d d for, simply reducee bth y d enrichment. Increase abouf so 0 gram5 t 235f elemenr so U pe t result in no net reactivity change when the fuel element geometries are the same. Reactivity effects due to fuel element geometry differences are slowly varying function numbee th f platef ro so elementr pe s fuee th ,l meat thickness, and the water channel thickness.

REFERENCES

. ZolotarA . B 1 .al.t ,e , "EPRI-CELL Code Description," Advanced Recycle Methodology Program System Documentation, Part II, Chapter 5 (1975).

2. J. T. Thomas, "Nuclear Criticality Assessment of Oak Ridge Research Fuel Element Storage," ORNL/CSD/TM-58 Ridgk ,Oa e National Laboratory (1978).

3. E. B. Johnson and R. K. Reedy, Jr., "Critical Experiments with SPERT-D Fuel Elements," ORNL/TM-1207, Oak Ridge National Laboratory (1965).

Derstine. L . K ,. 4 "DIF3D CodA : Solv o et e One-, Two- Three-Dimensionad ,an l Finite-Difference Diffusion Theory Problems," ANL-82-64, Argonne National Laboratory (1984).

5. L. M. Pétrie and N. F. Cross, "KENO-IV: An Improved Monte Carlo Criti- cality Program," ORNL-4938 Ridgk ,Oa e National Laboratory (1975).

6. R. N. Blomquist, et al., "VIM - A Continuous Energy Monte Carlo Code at ANL, M ORNL/RSIC-44 Ridgk ,Oa e National Laboratory (1980).

. GilleyW . L d , an "Critica x Fo . K l . ExperimentJ . 7 s witd han ArrayR OR f o s BSR Fuel Elements," Neutron Physics Division Annual Progress Reporr tfo Period Ending Septembe 1958, r1 , ORNL-2609 Ridgk ,Oa e National Laboratory (1958).

127 APPENDIXA

STORAGE RACK CALCULATION FUNCTIONS SA FUEF SO L ELEMENT ORIENTATIONS. WATER REFLECTOR THICKNESSES. AND FUEL ELEMENT SPACINGS

FUEL ELEMENT ORIENTATION

Infinite Row of Fuel Elements

Eigenvalue calculations were performed using three dimensional (XYZ) diffusion infinitn theor a fuef o r lw yfo elementero functioa s e sa th f no separation between fuel element elemene th d tsan orientatio fuee th l n ni storage compartments. Three orientations of a fuel element in a storage compartment were con- fuee th l f sideredelementi s a firse Th s .st wa were placed wit fuee hth l plates perpendendicular to the row; the second, as if the fuel elements were square (no directional distinction); and the third, as if the fuel elements were placed witfuee th hl plates paralle directioe th rowe o th lt .f no Figure Al depict these three fuel element orientation schemes in an infinite row of fuel elements.

REFLECTOR

1 FUEL PLATES PERPENDICULAR ( i. I TO »01 _ _ INFINITELY LON - G«ID c. B 3 YE1 FUEL REGION '(33 | 5> DEFINED ZERO SEPARATION (S)BEIIEEN fUEl ELEUENTS H« -0 0 c. BETWEEN FUEL REGIONS 1 \ t -4k ^SDE PUTE REGION ' REFLECTOR

SQUARE REPRESENTATION OF FUtL tLEUtHT t _ _ INFINITEL FUEM H L REGIO- Ya LON1 1 NÏ GB FUEL REGION 7.1 (H>9> DEFINED ZERO SEPARATION (S) BtTIEEN FUEL . 8£T»EEc ELEMENT 1 ~0 S N SFUEH» L RECIONS 1 •41- REFLECTOR ,-FUELPLMEREClOM " P ' n FUEL PLATES PARALLEt T1 ORO U I I • 1 INFINITELY LONG BÏ101.- »I0t FUEL REGION , to DEFINED ZERO SERRATION IS) 8CIIEEN FUEL 1 ELEUENTS KM - M » 8ET1EEN FUEL REGIONS

-^ si REFLECTOR

Fig. PossiblAl . e Fuel Element Orientation Rowa n i s.

The calculations were performem separatioc 0 2. do t nove 0 ranga r0. f o e between fuel element watem c d 0 wit1 ran s a reflectoh n eaco r e h th sid f o e rown thesI . e calculations e aluminiu storage th , th f o me rack compartments was neglected and only the most reactive LEU fuel element type was used. In this case, that would be the LEU U3Si 621 g U element (No. 13 in Table 1) since this element contain largese th s t fissile loading e end-fittingTh . n so e fueth l elements were also'neglecte calculationse th n i d .

128 Tabl . EigenvalueAl e Calculation Infinitn a r fo s e Row Results of of Number 13 Fuel Elements (LEU t^Si 621 g these diffusion Functioa s a 23 ) 5f Elemenu no t Separation ni theory calculations a Row and Element Orientation in a Row. are shown in the (See Figs. A, B, and C in Appendix C.) first part of e th n i d Tablan l eA Fuel Square Fuel bottom three curves Element Water Element Plates Representation Element Plates Separation Reflector Perpendicular of the Parallel of Fig. ClearlyA2 . , In a Row, cm Thicknessm c , to the Row Fuel Element to the Row e fueth l element orientation in a 0.0 10.0 0.7920 0.8038 0.8154 reactiva s rowha - 0.25 10.0 0.7937 0.8049 0.8161 ity effect. When 0.50 10.0 0.7942 0.8046 0.8153 the fuel plates are 0.75 10.0 0.7934 0.8031 0.8132 parallel to the row, 1.00 10.0 0.7914 0.8002 0.8098 this results in the 1.50 10.0 0.7838 0.7910 0.7994 largest amounf o t 2.00 10.0 0.7718 0.7775 0.7849 fuer unipe lt length of the row and therefore, is the most re- a C 0.75 10.0 0.8250 0.8311 0.8458 active orientation. 0.75b 7.0 0.8912 0.8965 0.9141= e Fothreth r e _ _ orientations of fuel 0.25 7.0 0.7909 elements in a row, 0.25 8.0 - - 0.8036 the eigenvalues 0.25 12.0 - - 0.820Z .3 62 o t diffep u y b r 1.77 8.6 - - 0.7860d 6k/k for zero separ- tion between fuel aData for an lnflnlte-by-lnfinlte array with a 20-cm row separation. element 1.3d an s2 n infinlte-by-lnfinlta *>Dat r fo a e array wit a h14—c separationw mro . 6k/k* for 2 cm sepa- cSee Table 2 for parallel-place data. e rationth s A . d keff Is 0.8431 for an inflnlte-by-infInlte array with a 17.2-cm row separation. separation increases, e fueth l element orientation reactivity effect is smaller.

s alsi t oI evident thamose th t t reactive configuratio t zera t o no sep s i n - aration between fuel elements; the optimum separation is between 0.25 and 0.5 cm depending upon the fuel element orientation. This suggests that for separations larger than the optimum separation, water effectively begins to isolate a fuel element from its neighbor. Source multiplication experiments3»7 suggest that about 17 fuel elements, with optimum spacing, are equivalent to an Infinite row of fuel elements.

Inf tnite-by-Infinite Array of Fuel Elements

Fuel element orientation reactivity effects were also examine r configufo d - rations with more than Jus singla t e infinit f fueo w l ro eelements . Calculations were performe r infinite-by-infinltfo d e array f fueo s l elementd s an wit 4 1 h 20 cm of water between rows. These result e secone showth ar s n e plottedi n ar par d f Tabln o an i td l eA same Th e . generaFigA2 . le eigenvalutrenth f do e bein a gfunctio e fueth lf o n element orientation in a row is evident in these data, as is also, the effec- tivenes f wateo s n isolatin i rfro w ma ro neighborin e separatioe th gon s a w gro n between rows increase.

*For convenienc n thii e s report e reactivitth , y unit 6k/ s k definei 6ke b effo t d.

129 e optimuTh m fuel element orientatio n thesni e types of storage configurations si therefore, wit fuee hth l plates of the fuel element paralle rowse th .o t l This orientation gives the maximum eigenvalu d thereforeean s ,i the most conservative in judg- ing the acceptable reactivity fo storaga r e rack configuration.

WATER REFLECTOR THICKNESS

Infinite Row of Fuel Elements The reactivity effect of

» I I JO various thicknesse watef so r CXCttEMT SEPARATION <* on an infinite row of fuel elements was examined to de- Fig. A2, Eigenvalues for Various Fuel Element termine the effective infinite Orientation n Infiniti s e Arrayf o s thicknes reflectof so r mater- LE Ug 235 U1 3uS62 i Fuel Elements ial. These calculations were as a Function of the Separation performe storaga r fo d e con- Between Elements. figuration with optimum spac- UU 3SLE i inf o g ) (0.2cm 5 621 g 235U fuel elements and 7, 8, 10, and 12 cm water re- flectors. As in the fuel

>,„ n O.BIEIII element orientation studies, MI« u K nMcrm or MUAT rrrc Ml PUTES PAHAU.ll. TO K» the storage rack compartments and the fuel element end- fittings were neglected. iK DATA FOR- t scMMTito K mr o These results are listed in the third part of Table Al and are plotted at the bottom of Fig. A3. These data show that the reactivity increases by just over 1% 6k/k as the reflector thickness increases increased an , cm s8 o frot m7 only another 1% 6k/k for a 2 cm increase from 8 to 10 cm. l(e»o The Increase in keff from 10 to 12 cm is substantially smaller which indicates that "» I—I—l—l—I—l—l—L l i i i l i i i l i i i watef o nearls m 1ri c 2 n a y » ! » ! » l « l « • • infinite reflector with res- HUMI SCPMATIW « n infinita pec f o o t w ero Fig. EigenvalueA3 . Variour sfo s Reflector fuel elements. Thicknesse Infinitn si e Arrayf so LEU 235g U 31 Sui 62 Fuel Elements aFunctioa s Separatioe th f no n Between Elements.

130 Inflnite-by-Infinite Array of Fuel Elements

As an extension of the above data, the reactivity effects of adding additional rows of fuel elements were examined. These data are also plotted in Fig. A3 for 14, 16, and 20 cm thicknesses of water between rows of an infinite-by- infinite array of fuel elements. These results indicate thacoupline tth g between rows rapidily decreases sa the water reflector thickness increases. The reactivity difference between a single infinite row of fuel elements with a 10 cm reflector and an infinlte-by- infinite arra watef o ym rc wit betwee0 h2 n row onls si y 3.4Z 6k/k. This reactivity difference Increasewatem c - 4 1 rre d 12.9%o st an cas7 e f th 6k/eo r kfo flector thicknesses.

Afunctiosa separatioe th f no n between element rowa n si , this coupling decreases as the separation increases. For example, with reflector thicknesses coupline th , 8.3s cm i g 6 %1 6k/d an 0.2 okt f8 a separationm 5c 5.7d ,an 2 fik/k at 1.77 cm separation with reflector thicknesses of 8.6 and 17.2 cm.

FUEL ELEMENT SPACING 1.030 A resula sindin a f -to cated optimum fuel element spacinn effectiva d an g e

',„ n ElUEKT SEPARATUM infinite reflector thick- 001 ness for a single row of 1.000 _ DAT» FOR »FllllTE-BY-l«FmiT£ ARRAY FUHCT10A l S A S(M F KO SEPARATION, o I SEPARATION. fuel elements n examin,a - FUEL PLATES PARALLEL TO ROK. ation was made to determine the corresponding optimum spacin d reflectoan g r thicknes n infinita r sfo e 0.950 arra fuef o y l elements.

The calculations assumed an infinite—by-infinite g U 3arra1 SLE i62 f yo 235ij fuel elements witw ro h 0.900 separation o t frof s o 0 1 m d elemenan , 22cm t separa- n I . cm 0 2. o t 0 tion0. f so all cases, both the alu- storage minuth f o me rack and the fuel element end- fittings were neglected.

e resultTh f thesso e calculations are shown in Table 2 and are plotted in Figs. A4 and A5. The data 1.00 1.50 2.SO ELEMENT SEPAMTHMm c . are plottea Fign s i a d 4 .A function of element separa- . EigenvalueFigA4 . r Varioufo s Separation w Ro s s tiovariour nfo w separasro - n Infinite-by-Infiniti e Array f LEo s U a s Fign tionsa i 5 d .A an , 1738! 621 g 235[j Fuel Elements as a function of row separation Function of the Separation Between for various element separa- Elements. tions.

131 Optimum Fuel Element Spacing

As clearly shown in Fig. A5, the optimum fuel element spacing is a function of the row separation. For row separations greater than about 18 cm, the optimum fuel element spacing is about 0.25 cm. This optimum spacing is about th s enote wa sam ds ea previousl infinitn a fuef o r lw fo y ro eelement s witha 10 cm reflector.

Effective Infinite Separation

1050

In case of coupling between rows, Fig show5 .A s the slopes of the 1000 — DATA FOR INFINITE-BY-INFINITE curves are becoming ARRAY A FUNCTIO S A S F ELEMENNO T SEPARATION fairly flat when FUEL PLATES PARALLEL TO RO»S w separatiothro e n is about 22 cm. For larger separations reactive ,th - 0950 — y effecit t wile lb substantially smaller which indicates that 22 cm separation between rows is nearly an infinite separation. 0900 — This separation is consistent with the m 1reflectoc 2 r thickness noted previously for an infi- fuef o nit lw ero ELEMENT SEPARATION. Cff elements. 0850 —

200 _ 0600 14 !( » SEPARATIORQ N

Fig. A5. Eigenvalues for Various Element Separations In Infinite-by-Infinite ArrayU U$SiLE f so 621 g 235U Fuel Elements as a Function of the Separation Between Rows.

132 APPENDIXB

EIGENVALUE VARIOUR SFO S FUEL ELEMENT TYPES IN VARIOUS STORAGE RACK CONFIGURATIONS

CHANGING THE STORAGE RACK CONFIGURATION storage Th e rack configuration considered R herOR simulas e i th o t r storage rack n thiI . s case same ,th e 1.7 spacinm 7c g between fuel elements is assumedseparatioe th t ,bu n between row decreases si d fro. mcm 17.4 1 o 2t

Based upo date reactivite th nf Figth ao , .2 y chang r thiw confiefo s ne - guration is about +4.4% fik/k. When this 0.044 change in eigenvalue is added to the curves of Fig. 3, the estimated eigenvalues for the various LEU and HEU fuel element type e obtainedsar ; these parallel curve e showsar Fign ni . .Bl

As a check of these

111 Mil estimates, eigenvalues j iQl ' I ' 1 'L were calculated for five FUEL STORAGE RACK tlTH 1.766 CH/ ELEMENT of the twenty fuel ele- SEPARATION AND 14cal ment types in this new ROI SEPARATION configuration. These data point listee sar d in Table Bl and plotted i nthesn I Fig e. .Bl calculations, both the aluminum storage rack compartments and the fuel element end-fittings were included.

The data in Fig. Bl show thae threth tU eLE and the two HEU fuel types very nearly fal thein lo r estimated eigenvalue UAI PLATEU 3 1 ,HE SV curves. Based upon this UAI U PLATE3 7 ,HE SA Indicated good agreement, an Important observation •jo D «-EU UjSi IS PLATES UjS U PLATED i LE SO can be made. This is, that eigenvalue scaling g U UßS 1 LE usin i62 e th g 235U fuel element data (Fig. 2) works well for 1.1$ Turn min in 111 il 111 llllll other fuel type d 235jsan j m o so o «o 0 30 HO m loadings cant I . , there- •"U MASS/ELEMENTG . fore, be expected that . FigBl . EigenvalueU HE r Varioud fo s an U LE s reasonable eigenvalue Fuel Elements in Infinite-by-Infinite estimate made r b fo en sca Arrays With Separation f 1.76o s m 6c wida d an e U rangLE f eo Between Elements and 14 cm Between HEU fuel element types Rows as a Function of the 235U Fuel ivarieta n f storago y e Element Loading. rack configurations.

133 Table Bl. Eigenvalue Calculations for Five Fuel Element Loadings, Each Infinite-by-Infinitn ia n e Array Assumin Elemenn a g t Separation of 1.766 cm and a Row Separation of 14 cm. (See Fig. F in Appendi) xC.

235 Fuel Element Fuel Plates/ U7 keff Loading Number Type Element3 Element Eigenvalue

9b U UßSLE i 19 395 0.8528

12C LEU U3Si 23 395 0.8394

13 LEU U3Si 23 621 0.8962

14 HEU UA1X 19 180 0.7300

18 HEU UA1X 23 280 0.8094

aEnd-fitting Al/^O volume fractions are: 19-plate element 38.98/61.0d 2an 23-plate element 44.53/55.47. Cranium density is 5.391 g/cm3. cUranium density is 4.454 g/cm3.

EXCHANGING HEU AND LEU FUEL ELEMENT TYPES

Based upon the data of Fig. 3, reactivity estimates can be made for fuel element geometry differenced Usan loading changes givea r nFo . U fuel element loading, these reactivit 235 y estimates ar235e summarized below:

Fuel Element Configuration Change Reactivity Change same th e U fue r HE fuelfo . l vs elemenU LE 1. t 6k/% 3 o kt les 2 s geometry reactive 2. 235U loading increment for the same fuel element 0.5X 0 6k/1 r kpe geomy etr grams

3. Number of plates per element for the same fuel -0.5Z 6k/k per meat thickness plate

4. Fuel meat thickness increment for the same number -2.0% 6k/r pe k of plate elemenr pe s t 0.25 mm

134 As an example of how these reactivity effect figures can be used, the estimated eigenvalue 235g fueU U0 HE l 30 elementgeometrier o sfo tw d san f so 235g fueU U 0 LE lstoragR 40 elementOR e th e n racsi k configuratio givee nar n below:

keff ORR Storage Rack Configuration Loading Estimated Figure3

— — 235g U0.790 platesHE9 ,1 30 U ,1 0.5 mea1m m t

LEU 400 g 235U, 23 plates, 0.51 mm meat 0.796 0.793

1. LEU vs. HEU: 0.791 - 0.025 = + 0.766

2. 235U loading0.00* 0.05+ 0 1 5= + 0:

3. Numbe platesf 0.00* o r 0.02 - 4 5= - :0

4. Fuel meat thickness: no change

235g U0 platesLE2 ,2 40 U , 0.7 meam 6m t 0.781 0.781

HEU. vs :U 0.79LE 1. 0.021- 0.76+ 5= 6

2. 235U loading: + 10 * 0.005 = + 0.050

3. Numbe platesf 0.00* ro 0.01- 3 5= - :5

4. Fuel meat thickness: - 1 * 0.02 = - 0.020

In this particular example there would be very little change expected in the storage rack reactivity with the three fuel element types.

Whil above th e e reactivity coefficient e approximatar s d applicablean e ove limitera d range, they provide base estimato t s e t whane te woulth e b d effec storaga n o t eU fue LE rac lf i kelement s wer replaco et U fueeHE l elements.

135 APPENDIXC

THREE-DIMENSIONAL DIFFUSION THEORY CALCULATIONAL MODELS

Y (A) Y (B) Y (C) D*'-K).47*-0 Di).'-f0.47»-0 11-0(7) 13.8(10) ———————— | 13.9(10) ————————— , , to— — _ 16.0(12) 1

H20 H 2° -J H2° ! 1 4.00(1) 1 3.80(1) —————— l 3.90(1) Side i (•% Side * ' j All dimensions in cm. 3.15(3) — s 1 Fuel Region mesh points in IHM i 1 parentheses. A . ,,__.., ,,y\l.,...» " u*1 Z-dimension 10(4)e sar , 20(3), 30(4), 34(5), 45(3).

Fuel element half-height endn e wit-m 4 c a h0 i 3 s (D) Y (E) Y i (F ) fitting and an 11 cm axial 9.00(5) water reflector. to 15.0(11) Reflective boundary con- 12.62(8) dition (*'-0) on all 11.0(7) faces except where indi- »2° cated Z-45cd ,an m where H20 H20 Fuel element orientation 4.68(1) 4.68(1) Al Al in a row is indicated by 4.37(1) 4.37(1) fuel plates being either 4.00(1) 4.00(1) 4.00(1) 3.98(4) 3.98(4) 3.98(4) perpendicular to or parallel to the X-axis, ———— OJ O ——— -So TD

^~•N f •N l*~ ( m— — « - X C•> ^H .— - X

Y (G> Y (H) Y (I) Y (J)

t *> /f11 l \ 12.62(8)

11.0(7) 11.0(7) H20 2

H20 H20 4.68(1) 4.68(1) 4.53(1) AI Al Al 4.37(1) 4.37(1) 4.25(1) 4.22(1) 4.00(1) 3.80(1) 4.00(1) Al 4.00(1) 3.93(4) Side 3.98(4) 3.98(4) 3.15(3) o «ts, w ——— O . ——— . •o X . ——— •o •o r X inn X ^ s*. 5 -^^«. » r• / •\ <•î J ~ X C C- X C 7 X f>•) -^Cn ~ X

136 Appendix N-3.2

FRESH FUEL STORAGE

Y. NAKAGOME . KANDAK , . SHIBATT , A Research Reactor Institute, Kyoto University, Osaka, Japan

Abstract A criticality analysis on the storage of MTR-type medium-enriched uranium (MEU) fuel elements was performed. It was assumed that the MEU fuel elements were arranged vertically in two rows in a stainless steel box, and that the boxes were placed in parallel infinitely at a certain distance. In both sides of the box, 0.3w/o boron-loaded stainless steel plates were used. Moreover, it was assumed thae x insidwerth bo td eoutside an efilleth f p wito eu d h water. K f, was calculated as a function of the distance betwee e boxetn ny usin b s g ANISN-JR codee effecTh .f o t s alswa o, , examined boroK n o n A simila. r analysie th n so storage of HEU fuel elements was performed and compared with the results on MEU fuels.

INTRODUCTION

n fueO l management s importani t i , o stort t e fuel elements safeld an y efficiently in a fuel storage facility. When a great number of fuel elements are stored in the facility, safe and simple handling of fuels, radiation shielding, decay heat remova d criticalitan l y safety contro n appropriata s wela ls a l e physical protection are required. Especially, the problem of criticality safety contro peculias i l nucleao t r r material storage. At the Research Reactor Institute of Kyoto University (KURRI), many MTR-type fuel elements containing highly-enriched uranium (HEU, 93%) have been storeo t d be kept K ,, S 0.95 in a fresh fuel storage room and a spent fuel pond.

Recently e Reduceth , d Enrichmen r Researcfo t d Tesan ht Reactors (RERTR) program has been forwarded, and the KURRI is cooperating vith the Argonne Nation- l Laboratora e jointh tn e RERT i yth stud Rf o yprogram . Accompanyin- re e th g ductio f enrichmento n s necessari t i , o confirt y e criticalitth m ye safetth r fo y storage of reduced enrichment uranium fuels. This paper presents the results of criticality analysis on the storage of medium-enriched uranium (MEU, 45%) fuel elements.

MEU FUEL ELEMENT AND STORAGE METHOD

U fueME l e elemenTh t considere criticalite th n i d y analysi shows i s Fign i n . e dimensioTh 1 .elemene fuee e sam th th s th tha l a e f f s no elemeno ti t f Kyoto t o University Reactor (KUR). One fuel element is composed of 18 fuel plates. The clearance (water gap) betwee platee th n 2.8s i s. Eac1mm h fuel plate consistf so a 45% enriched uranium-aluminide (UA1 -Al) meat cladded with aluminum. The thickness of the meat is 0.5 mm and that of the clad is 0.51 mm. The average length of the meat is 59.4 cm. The.uranium density of the meat is 1.6833 g/cm U-23e anth d 5 densit 0.757s i y 5 g/c. m

137 The storage method for the MEU fuel elements is also same aa that for HEU fuel element KURf o s e fue. th l Tha , elementis t e arrangear s d verticallo tw n i y rows and stored in a rectangular box made of stainless steel. The fuel storage box is shown in Fig. 2. In both larger sides of the box, stainless steel plates containing natural boro e partiallar n y usede positio e boron-loadeTh .th f o n d stainless steel plates corresponds to the meat position of the fuel element. The content of natural boron is about 0.3w/o, and the thickness of the side plate is e fueTh l . storagmm 2 e boxe e placesar n parallei d a suitabl t a l e distanc. a n i e storage e buildin rood th e fixestorag an th mo t dn gi e floor, roo, ë i K m. controlled less than 0.9 5rooe s th filleevei m f i nwitp u d h water.

- A 61. 11 -JL B r'

C\\^sT bÂA-.., ..,.- r i-=s \f~~^- jijij'._ ii -.—".-.•:. ^'±.Â.'aîx4rai&iiKur£s*za*i i — • • ••- ' •-...• ..— .. , ji ^_^ , , ——— —— — 127 ——— — - 69.85 büb. 10 •' O7-J ————————————————0 1 ....

-75.40

79.1

. Fig1 MTR-typ. e MEU fuel element considere n criticaliti d y analysis (scale : mm)

r-» B .1 '^u

-i 90 - 90 r 1 195 * H _A A A , . —— . 1175 ————————————— J

Side Plates 928 480 HTR type fue-l Element

A-A B-B

Fig. 2 Fue. l storag x (scal) bo e mm : e

138 CRITICALITY ANALYSIS

n ordeI calculato t r e storagU fueeth ME K, l r f fo eelementsfo i e dimen,on - sional transport code ANISN-^IT s usede numbewa r Th .f employeo r d energy groups therma5 fas1 1 1 , td 26 neutrolan s wa n energy groupscrose th sd ,an section s were taken from MGCL 26 library2.

The calculation was performed on condition that: (1 )e fueth l s fillestoragwa x d bo ewit h fresh MTR-type MEU fuel elements mentioned above, (2 )e insidth outsidd e box'weran eth f o e e fille witp u d h water, (3) for the region where the fuel elements were in line, macro cross sections were obtained by cell calculation so called unit cell - super cell, (A) the fuel regio consideres nwa a 62.5 s a d5 cm-high, infinitely long slab fuel, and (5) two slab fuels were combined with boron-loaded and ordinary stainless steel plate placed an s parallen i d l infinitel certaia t a y n distance. The above conditions are illustrated in Fig. 3.

Fuel plate

Q) U-A1

(V) Fuel legion

Fuel element

SUPER CELL

Side plate \ Ordinary Stainless Steel or 0.3S boron- loaded Stainless Steel

Center plate "~ Fuel Region

H20 Stainless Steel Side plate

Fig. 3. Outline of criticality analysis.

139 , calculation, K e th Prio e validito th t ,r f usino y g ANISN-JR cods wa e confirmed by comparing the experimental K ...-value with the calculated value. e experimentaTh l valu s obtaineewa d frocriticae th m l experiments usin U fueME gl in the Kyoto University Critical Assembly (KUCA). As a result the calculated K .,-valu n agreemeni e eb o turnet t t witexperimentaou e d th h l value withi. 1% n

K ,, was calculated as a function of the distance between the fuel storage boxes n ordeI .o examin t r e effecth e f boroo t n containin e sidth e n plati g n o e e alsw o, f , Kcarrie t similaou d r calculatio e cas f th usino e K ... r .gfo f o n ordinary stainless steel side plates. Moreover, a series calculation of K ,, was performe e storagth r f MTR-typo efo d U fueHE e l elements e resultth d san , were compared wit U fue resulte U fuelsHE th h lME e propertie d meatf Th an s.o U sME f so calculatioe th use n i d e listenar n Tabli d . 1 e

Tabl fuelU . HE 1 e Properties d usean criticalitn U i d ME f so y analysis.

MEU HEU

Fuel material UAlx - Al U-A1 Enrichment (%) 45 93.15

Uranium loading (%) 42 20

Meat density (g/cm) 4.0081 3.2598

Uranium density (g/cm ) 1.6833 0.6520

U-235 density (g/cm3) 0.7575 0.6073

RESULT DISCUSSIOD SAN N

The calculated results of K ff are shown in Fig. 4. In the figure, 'without-boron r 'with-borono ' mean a scas f usino e g ordinar r 0.3o y % boron-loaded stainless 1 steel side plate, respectively. In the case of with-boron for MEU, the calculated K , is less than unity when the distance (D) between the fuel storage boxes is larger than 6 cm and less Chan 0.95 at D £ 8 cm. If ordinary stainless steel plates are used in both sides of the box, D à 10 cm Is required to be r IIE Fo less U i subcriticafuels less. i cm sf s K f 5 tha1 K nÄ d lan 0.9D t 5a than 0.95 at D 0 J cm and 10 cm in cases of with-boron and without-boron, respectively. These result summarizee sar Tabln i d . e2 n thiI s calculation becoms ha t ,ei clear , decreasetha, K t s abouo 10t % 6 t when 2-mm thick 0.32 boron-loaded stainless steel plates are used in both sides x insteabo e oordinarf th fo d y stainless steel plates.

s abous comparinA i % largeU b t ME U wite result r ME th g h fo r HEU, fo s, ,K than that for HEU in both cases of without- ana with-boron. This is mainly difference causeth U-23e y b dth 5f eo densit neate th .f o y

140 1.300 without boron — with boron 1.200 HEÜ •——- without bonon boron 1.100

1,000

0.900 -

0.800 -

0 5 W 0 3 20

Distance between fuel storage boxes (cm)

Fig. 4 Calculated-. K ,,-value a funstio ere s a rsth f o n distance between fuel storage boxes for d HEUMEan U .

Table 2. Minimum required distance between fuel storage boxes.

MEU HEU

with- boron 8 cm 5 cm

without- boron 15 cm 10 cm

CONCLUSION

A criticality analysi e storagth n f freso so e h MTR-typ U fueHE el elements s beeha n s assumeperformedi t calculatioe i th d, f n , thaI arrae .K th tf f o yo n the fuel element e considerear s n infinitela s a d y long slab fuel fuee d th ,an l its surroundings are filled up with water and the slab fuels are placed in parallel infinitely at a certain distance. A similar analysis for HEU fuels has also been performe same e compareth b e n o conditioo dt U d ME wit resulte f o nth h s From these results of the K ,, calculation, it has been concluded of MEU fuels, ef f that:

) sinc (1 U-23e th e U fue5 ME densitle considereth f o y n thii d s analysis is higher than that of the HEU fuel, it is required for the storage of MEU fuel element keeo distance t s th p e betwee fuee th nl storage boxes abou5 1. t times larger than the case of HEU fuel storage,

141 usiny b ) mm-Chicg2 (2 k stainless steel plates containing 0.3% natural boron i n both sides of the box instead of ordinary stainless steel plates, it is possiblU fueHE lr o elemento stor t eU amoune times2 ME th 1. e f y o t.sb

As mentioned above, it is valuable to use the boron-loaded stainless steel plate in the fuel storage box when a great number of fuel elements are stored in a limited area of the storage facility.

Since the fuel storage boxes of KURRI are placed in parallel at a distance of large randm c tha 0 ,6 n moreover, 0.3% boron-loaded stainless steel platee ar s used in both sides of the box, our storage method is sufficient for the storage fueU ofME l element criticalite th n i s y safety control with considerable surplus.

These results should also be useful for spent fuel storage in a pond.

ACKNOWLEDGEMENTS

e authorTh e mosar s t gratefu Messrso t l . Hayash. M ShircyS d f KURRan io a I for their helpful advices and operating computers.

REFERENCES

Koyama. 1.K Taji. ,Y Minami. ,K Tsutsui. ,T Miyasaka. . IkedS ,T d aan A ," one-dimensional discrete ordinates code for neutron and gamma-ray transport calculations", JAERI-M 6954 (1977)

Naito. Y . 2 Tsuruta. ,S . Matsumur,T Ohuchi. T d an a , "MGCL-Processor ;ComputeA r Code Syste Processinr fo m g Multi Group Constants Library MGCL", JAERI-M 9396 (1981)

142 Appendix N-3.3

SPENT FUEL STORAGE*

B.C. MERELLES TA . A , Netherlands Energy Research Foundation, Petten, Netherlands

Abstract

A criticalitY„analysis on the.storage of highly enriched HFR fuel elements, side th e n platesi B d 100 performeds an , 0g wa U m containin e fueg . Th 0 l g42 element pooe storee specialln th sar i l n i d y designed compact racks, consistinf go cladded cadmium boxes. For reasons of safety and flexibility the analysis of the (spent) fuel elements was performed for an infinite array of fresh elements with 450 g U without UB in the side plates. In the framework of the safety and licensing guidebook additional calculations were performed for LEU fuel elements, containing 450, 475, 600, 675 and 1000 g U respectivel d compareyan fueU d HE witl resulte e elementshth th f so . Alse oth reactivity effect of the cadmium boxes and the presence of Be elements in the boxes was examined.

INTRODUCTION

Up till 198 storage licence 2th th r fresf o efo h fuel element pooe th l n si side th ewitU s restricte n platesi wa h B g element1002 1 t 5 .I g 0o m 40 t d f so was not allowed to store both fuel and beryllium elements in the racks. The old storage racks consistex position si rowo f tw so f o ds separate thica y db k layef ro water and at the outer sides provided with a cadmium liner with a height of 30 cm. In order to store fuel elements more efficiently and to get rid of the above mentioned restrictions new fuel racks were designed for under water storage of (spent) fue othed lan r elements suc beryllius ha m reflector elements poow ne l e .Th storage racks are provided with a 1 mm thick cadmium box around each element position Fige numbeA ,se . 1 f thesro e boxe placee sar d nex eaco t h othea n i r tank whic uppee opeth s hi r t na side . r reasonFo safetf so flexibilitd yan analysie (spente th y th f so ) fuel _„ elements.was performed for an infinite array of fyftsh fuel elements with 450 g U without R in the side plates. An analysis with B in the side plates is more complicated due to the.reactivity increase,.during the first burn-up steps caused by the fast depletion of B with regard to U. Also the reactivity effect of replacing fuel elements by Be elements was examined. In the framework of the safety and licensing guidebook additional calculations were performed for LEU elements, containing 450, 475, 600, 675 and 1000 g U respectively. These results were compare dU fue wit HE resulte le th helementsth f so . Also the reactivity effects of the cadmium boxes were examined both for LEU and HEU elements.

wore Th k* describe thin di s repor bees tha n carrie undet dou r contrac Europeae th o t n Commissiod nan has bee budgetC n JR finance e .th y db

143 82.0 (mm)

Materials ; k » Aluminium B • Cadmium C • H20 0 • Fuel element

Fig. 1. Horizontal cross section, of a 23 plates fuel element in a storage box, provided with cadmium plates in the four aluminum walls.

CALCULATION PROCEDURE

The multiplication e fillefactoth f do r storage racks were computed wite th h aid of the 2 dimensional diffusion code TEDDI-M |3|. The nuclear constants, require n thii d s cod e computear e r fivfo d e energy groups. These energy groups were :, , - 1.35eV V e 0 31 0 1 4 1 Grou = 1 p - V Grou0.067 e 1.35 V = ° ,e 2 p 10 4 0 3 1 Group 3 = 0.0674 10 eV - 0.683 eV 3 Grou0. 0.68 = - 4 pV 3e 0 - V e 3 Grou0. = 5 p thermal groups

144 e computeTh r program e determinatiosth user fo dnucleae th f no r constants were : epithennae th r GGC-I) fo a d fas\ lan \2 Vt groups (above 0.68) 3eV thermaMICROFLUX-e ) b th r lfo group| |l 2 s (below 0.683 eV). The MICROFLUX calculations wer followine eth firs : carrien i y tt gwa deterou d - mination of the "flux weighed" number densities in the fuel region, consisting of the fuel O meatcoolinH claddine e ,th th gd channean g l (see Fig . Thes2) . e "flux weighed" number densities were used in the fuel region (material 1 in Fig. 3) of the second microflux calculation of the whole box.

MAT. 1 2 3 MAT.1 (fluxweight«dconcentr.) 2 3 4 3 3

0.0396 0.2 O. 0,025£ 0.1925 1 O. 7" 0.1109 (cm) _^ 3.1575 ———— \ O.45_ 0.3 f7- (flux weight. concen ) 1

A B <-j — ' n -1 TEDDI MATERIALS ' mat. 1 = fuel mat. 1 = flw. cone, of fuel, Al and H„0 mataluminiu= 2 . m l H= matA 2 02 + . 0 H = mat 3 . mat. 3 = Al matH2<= 4 ). mat. 5 = cadmium

Fig. 2 Geometry used in first Fig. 3 Geometry used in second microflux calculation. microflux calculation.

At the boundaries of the cadmium material region the so called black boundary conditio o neutron(n n s returning frocadmiue th m m material e applied th )ha r fo d lowest thermal group numbe. r5

The multiplication factors were calculate n infinita r dimensionao fo d tw e l array of storage racks, placed next to each other. So at the four outer boundaries of a single box the zero next current boundary condition had applied. The buckling verticae factoth n ri l direction diffusioe th use, n i d,B n cod s 0.019ewa , corresponding with a fuel height to 60 cm and taking into account a reflector saving of about 8 cm.

CALCULATED CASE RESULTD SAN S

The criticality of the storage racks were determined for the following cases : Case 1 : The storage racks only filled with 450 grams U HEU fuel elements (94.4% enriched) o geometrica.Tw lcadmiue modelth witx r mbo fo sh fuel elements were used (compare Fig. 4 and Fig. 5 case 1). The multiplication factors, 0.55e 0.55d ar 5, an differenca 3( modele pcmk th 0 r 19 s )fo f giveeo n in Fig. 4 and Fig. 5 (case 1) respectively.

145 Case 2 : The storage racks partly filled with Be elements in the ratio of three fuel elements against one Be element (see Fig. 5). The k for this combination is 0.475. Case 3 : As case 2, but the numbers of fuel and Be elements are equal (see Fig. 5), The k ,. in this case is 0.370. err Cas numbee elemente B th 4 f t time3 ro numbebu s case s A si , th s 2 e fuef o r l element 0.239s i s. k (se e eTh Fig. 5) . Cas U elemente LE 5r fo Asst (20bu cas %, 1 e enriched) e storag.Th e racke sar completely filled with fuel elements containing 450, 475, 600, 675 and U respectively100 g 0 resulte .Th givee s ar Tabl n ni . 1 e Case 6 Imultiplicatioe n th additio , 5 d caso an t n 1 e n factorstorage th f so e racks without cadmium boxes were calculated for the case they were filled completel elementsU LE yr o wit e resultU .Th hHE alse sar o presenten i d Table 1.

COL UMN -»- 1! 2 <) ; 4 9 141 16 ) materials C(4X)| D ———————— i — 2 A fuel element B (4x) B H20 A C AI + Cd D Al

Distances in cm

9 Column : = 16-1 5 80. = 2 - 0 = 15-1 3 60. = 2 -3 0.25= 3 -4 = 14-15 1 4-17. = 4

Row 0- 2 = 0.5 = 16-18 ••M 2- 4 = 0.55= 14-16 9 4-17. = 4 1 I

Fig. 4 Geometrical model used in the first TEDDI-M calculation for storage racks, filled with fuel elements only.

146 A ( B Ei A A A

( £ -GB- C -FV- •

A ( 3 E> A A E

Case 1 : 4 fuel elements (A) Case 2 : 3 fuel elements (A) k ,.. = 0.553 ef f elemene B 1 ) + (E t 0^47- , , k5 eff

Material : A »elemene B fue = lE t element d C + I A = C

0 2 H = B D T AI

E A E E

-P_l- -Pa-

A E A E

Case 3 : 2 fuel elements (A) Case 4 : 1 fuel element (A) 2 Be elements (E) elemente B 3 ) (E s 0.37= , , kC k ,, - 0.239 ef f eff

Fig Geometrie. 5 . materiad san l composition TEDDI-e s th use n i dM calculations of the storage racks filled with fuel and Be elements.

147 Tabl Multiplicatio. 1 e n factocompacR HF f ro t storagU LE eU (90% d rackHE an )r sfo (20%) fuel elements.

HEU /LEU U-23, 5 ^ k ingra elemenm t with Cd withoud C t HEU 450 0.56 1.45 LEU 450 0.54 1.33 LEU 475 0.56 1.34 LEU 600 0.64 1.38 LEU 675 0.68 1.39 LEU 1000 0.82 1.41 LEU > 1500 — 1.45

CONCLUSIONS

e storagth f Accordino e analysese . racksth _, o k t g , e full,th y loaded with highly enriche 0 gram fueU 45 d sl elements withoulowes i , rB ttha n required. If these fuel storage rack e partlsar y fille witp u dreflectoe hB r elementse ,th multiplication factor wil evee lb n lower e samw .Th lo e use e r rackb fo dn sca enriched fuel with highe contentsU r . e compacth n I t storage racks whicn ,i h each fuel elemen place. ir t i i d ;.r> a cadmium box, there is only a small difference-in reactivity between FEU and LEU fuel elements containin same th gWithou. e U amoun cadmiue th tf o t m absorbing plates the difference in reactivity is significant.

REFERENCES

n Ligte . Tas va A d A.WlRoutinn |. ,an D nde . g MICROFLU calculatioe a progra th , X2 r e thermafo mth f o n l group constants, RCN-94.

Adir. J | ,2 S.S. Clark Fröhlic. ,R L.Jd han . Todt User programmerd san s GGC-I e manuath r V lfo multigrou p cross section code, G.A. 7157.

|3| J. Raven, G.J. Bakkenes, H.P. Struch TEDDI-M. Een programma voor de oplossing van de twee dimensionale neutronen- diffusie vergelijking voor veel groepen in x-y en r-z géométrie.

148 Appendix N-4

RECEIP FINANCIAD TAN L SETTLEMENT PROVISIONS FOR NUCLEAR RESEARCH REACTOR FUELS

UNITED STATES DEPARTMENT OF ENERGY Washington, D.C., United States of America

Abstract

U.S. Federal Registe 0 December3 noticef o s r (a s1987 n DOE')o s "Receipt and Financial Settlement Provisions for Nuclear Research Reactor Fuels e providedar " . DOE's current commitment to provide receipt and financial settlement services for quali- fying research reactor fuels extends unti 1 Decembe3 l r 198r 8fo HEU fuels and until 31 December 1992 for LEU fuels. The need for extension beyond these dates will be evaluated by DOE when appropriate.

149 Federal Tuesday/ Registe2 3 . ,No Vol / rFebruar, .51 198, 18 y 6 Notice/ s

Receip d Financiaan t l Settlement Provisions based upon the estimated actual cost of r Nucleafo r Research Reactor Fuels providing thi sE spenservicDO a t t fuea e l processing facility. AGENCY: Departmen f Energyto . In Its reviews of this policy extension, ACTION: Notice s determinedha o commerciaE N DO ) (1 : l fuel processing- ex U servicefuel LE e ar sr fo s e pecteavailablb o t d o meet e t anticipated SUMMARY e DepartmenTh : f o Energt s I y needs; (2) basic beneficial nuclear amending the provisions of its current research woul de curtaileb hav o t e d absent policy providing for the receipt and a spent fuel disposal capability; and (3) financial settlement of U.S.-origin spent U availabilitfuelLE n sa disposa f o y l research reactor fuel o Includt s e certain capability will encourag e conversioth e f o n research reactor fuels In which the research reactors currently using highly uranIum-235 content Is less than 20 percent enriched uraniu o t LEUm . s I Sinc t I e of the total uranium weight. Additionally, anticipated that these disposal services e Departmenth s providini t s financiaIt g l U fuelLE e b neede sr fo t wildno l until settlement terms for providing this completion of reactor conversions planned service. for the late 1980's, DOE proposes to extend the LEU fuel receipt provisions from the FOR FURTHER INFORMATION CONTACT: dat f thio e s publication through December Loui . WlllettR s , Mall Stop DP-131, Office , 1992 e 31 r neeextensioTh fo .d n beyond of Nuclear Materials Production. U.S. this time e wilb evaluatel d when Departmen f o Energyt , Washington, D.C. appropriate. 20545, 301/353-3968. To provide for Inclusion of LEU fuels In SUPPLEMENTARY INFORMATION n NovembeO : , 9 r DOE's current policy e term th d con, an s - 1982, the Department of Energy announced In E ditionserviceDO r fo ss describe paran I d - e Federath l Registe s rextendin wa tha t i t g graphs numbere e Federa th 1 throug dn i l1 1 h until Decembe , 1987e 31 r polic s th , it r fo y Register notice entitled "Receipt and receipt and financial settlement of U.S.- Financial Settlement Provision r Nucleafo s r orlgln spent research reactorR F fuel 7 (4 s Research Reactor Fuels" 47 FR 50737 pub- 50737) t thaA .t time e provisionth , f o s lished November 9, 1982, are hereby deleted this policy were restricted to: (1) and the following substituted In place: uranium-aluminum research reactor fuels with enrichments; I.e., uranlum-235 content 1. This policy applies to Irradiated as a percentage of total uranium weight, of nuclear research reactor fuels and blanket greater than 20 percent; and (2) uranium- materials {reactor materials). This policy zirconium hydride TRIGA fuel types. In pertains onlo t reactoy r materials from this notice, DOE Indicated that It was research reactors other than those Involved studying the reprocessing of uranium- In the conduct of research and development aluminum fuel compositions with uranium activities e leadindemonstratioth o t g f o n enrichment f o less s tha 0 2 percenn d an t the practical value of such reactor for would extend the provisions of the notice Industria r commerciao l l purposes. o Includt e thes U fuela reprocess LE ef I s - ing capability could be established. . Commercia2 l fuel processing muse b t unavailable at reasonable terms and s determineha E DO e d Th that reprocessing conditions. capabilitie U fuelLE sr fo wilse availb l - able and Is, therefore, prepared to extend 3. The fuel must be of U.S. origin - the provisions of its current policy to In- , thacomposeIs t f o nuclead r materials U uranium-aluminuLE clud e th e m fuel types produced or enriched In the United States. currently under development. The conditions governing DOE's offe o providt r e this . 4 This policy applies e solelth o t y service remain unchanged; I.e., that following types of reactor fuels: commercial fuel processing services muse b t unavailable at reasonable terms and condi- . Aluminum-claa d reactor fuels where eth tion d thae reactoan sth t r fuel f o mus e b t uranlum-235 content Is greater than 20 per- U.S. origin (composed of nuclear materials cent y weightb e ,tota th lf o , uranium con- produced or enriched In this country). The e tentactivTh . e fuel regio f theso n e fuel processing charges used for settlement e configureb fuel y ma s s uranium-aluminua d m under this program for LEU fuel receipt are

150 alloy, uranium oxidr o urantume - providing this service, In accordance with alumlnlde. Fuels containing significant the provisions of 10 CFR Part 1009. The quantitie f uranlum-23o s e excludear 3 d from baste charge r processinfo s g services will receipt. be reviewed periodical ly and adjusted when necessary. b. Aluminum-clad reactor fuels where the uranIum-235 conten s lesI t s tha r equao n l r thosFo e. 8 research reactor fuels to 20 percent by weight of the total describe n 4.aI dd 4.c .an . abovef o s a , uranium content e activTh . e fuel regions Januar , 19831 y e followin,th g charges will of these fuels may be configured as be applied to DOE processing services under uranlum-slIIcfde, uranlum-alumlnlde or th I s po11cy: uranium oxide. Fuels containing significant quantitie f o uranlum-23s e ar 3 r aluminum-claFo . a d research reactor excluded from receipt. fuels. Including alloy, oxid d alumfnldan e e composition, $1000 per kilogram of total c. Aluminum or stainless steel clad, delivered weight. Of this charge, $375 Is uranium-zirconium hydride (other than capital relate d $62 an s relatedI 5 o t d uranlum-233) TRIGA fuel types. operating costs; and

The percentage of uranIum-235 of the r aluminuFo . d stainlesb an m s steel-clad eligible fuel types shall be that measured uranium-zirconium hydride research reactor or estimated at the time of delivery to fuel, $105 r kilogra0pe f totao m l delivered DOE. weight. Of this charge, $395 Is capital related and $655 Is related to operating I undertakeI l w E DO , . 5 under contracts costs. Individually negotiated with persons licensed pursuant to sections 53.a.(4), The cap Ita I-related charges for DOE- 63.a.(4), 103 or 104 of the Atomic Energy provlded services shall be adjusted to f 1954o t d person,Ac an s operating research reflect changes In price levels from the reactors abroad fueled with materials base date of June 1982, In accordance with produced or enriched In the United States, the Official Monthly Construction Cost o posseswh r o wils l possess eligible Indices appearin n I "Engineering g News reactor materials, to receive such reactor Record. e operations-relateTh " d charges materials at DDE-designated facilities, and for DOE-provIded services shalI be adjusted to make a settlement, therefore. In to reflect changes In price levels from the accordance with this Notic d othean e r base date of June 1982, In accordance with establishe E policiesDO d . This settlement the Basic Inorganic Chemical Index appear- will take Into accoun e chargeth t r fo s ing In "Wholesale Price Indexes," published chemical processin f o receiveg d reactor U.Se bth y . Burea f Laboo u r Statistics. materials and any conversion of recovered e standarth uraniu o t md form, uranium 9. For those research reactor fuels hexafluorlde r whicfo , h specificationd an s described In 4.b. above, as of July 1, prices have been established by DOE. 1985, the following charges will be applied Furthermore y chemicallma E ,DO y procesd san to DOE processing services under this conver sucl al th received reactor materials pol Icy: e extentth o n t sucI , h manner t suca d h an , time, and place as It deems advisable, or a. For uranium oxide fuel compositions - otherwise dispos f suco e h materialt I s a s $660 per kilogram of total delivered y deema m advisable. weight.

6. DOE's commitment to provide fuel b. For uranlum-slI Ici de compositions - receipt and financial settlement services $835 per kilogram of total delivered will terminat e followinth n eo g dates: weight; and r researcFo . a h reactor fuels described r uranlum-alumlnldFo . c e composition- s In 4.a 4.c.-Decembed .an , 1987r31 d ;an $111r kilogrape 0 f o totam l delivered weight. . Fir7 m E servicechargeDO r fo s provided under this policy wile parb f l o t DOE will periodically review the charges each contract. These charge- sex wile b l for processin f theso g e research reactor presse a n termuniI df o ts weight charge fuels and revise said charges as fixed by DOE to the reactor materials In appropriate. question, to apply over the total number of units of weight. 10. The charge for conversion to uranium hexafluorld e purifieth f o e d nitrate salt e chargeTh r chemicafo s l processing of uranium that Is converted by DOE In Its services provided under this policy will processing of reactor materials Is $175 per reflect the Government's full cost for kilogra f containeo m d uranium.

151 11. A minimum charge of $44,500 will be applie o t eacd h batc f o fueh l material delivered to DOE under the provisions of Federa/ 2 l4 . RegisteNo , 51 Vol/ r . this policy. This charge reflecte th s Tuesday, Marc , 1984 h 6Notice/ s 7487 minimum cost to DOE of providing processing servicer fo small-batches d fuel Receipt and Financial Settlement Provisions materials e processin th e siz Th f o e. g for Nuclear Research Reactor Fuels batch to be shipped shall be as specified e persobth y n seekin e processinth g g services. DOE will permit a person to combine Correction Its batch with those of other persons In orde o t avoie rful th dl e th Impac f o t R DocF In . 86-3452 beginnin n o pagg e minimum charg r handlinfo e a smalg l batch e issuth 575 f Tuesdayn o eI 4 , Februar, 18 y size. Persons must notify DOE of their 1986, mak e followineth g correction: Intent to combine batches prior to the delivery reactoan f o ry materiale b o t s On page 5755, In the middle column, In Included In a proposed batch. Specific paragrap , subparagraph6 h s omittedwa b . arrangements must Includ a eformul r fo a Paragraph 6 Is corrected to read as distributing the processing charges and fol lows: other settlement factors associated with deliver e reactoth f o yr material o DOEt s. . 6 DOE's commitmen o t providt e fuel receipt and financial settlement services e optioth f compensatins o nha E DO . g12 will terminat e followinth n eo g dates: the reactor operator for enriched uranium recovered In the processing of reactor ma- r researcFo . a h reactor fuels described terials delivereE facilitieDO o t dn I s In 4.a 4.cd .an .Decembe- d an 1987, ;31 r accordance with the appropriate DOE-published price schedul r enrichefo e d uranium r researcFo . b h reactor fuels described material. Such compensation by DOE will In 4.b .Decembe- 1992, 31 r . consist of providing materials or services f equivaleno t value E willDO ., thereby, acquir ee uraniuth titl r o whict fo emt I h provides compensation E wilDO l . also acquire title, without cost, to all waste and other materials e containeth n I d reactor materials.

e enricheTh d uranium recovere n proi d- cessing reactor materials (or Its equivalent) delivered to DOE facilities and not compensated for by DOE, shalI be returned to the reactor operator. Enriched uranium will be returned to the reactor operator E processinDO f.o.b e th . g a site n I , reactor-operator furnished cask suitable for shipment offslte.

13. In lieu of processing uranium- zirconium hydride fuel types E wilDO , l agree to provide disposition services for such fuels thin I .s case o compensatio,n n for recovered uranium will be made. Research reactor operators may prefer to write off the value of uranium contained In e d fueaccepth an l t this service. Addi- tional Information concerning DOE's disposition service may be obtained from the Manager, Idaho Operations Office, U.S. Department of Energy, 785 DOE Place, Idaho Falls, Idaho 83402.

Dated: January 3,1986.

Sylvester R. Foley, Jr., Assistant Secretar Defensr fo y e Programs.

IFR Doc. 86-3452 Filed 2-14-86; 8:4] 5am BILLING CODE 6450-01-M

152 Federa lWednesday/ Registe0 25 . No Vol/ r, , Decembe52 . 198, 30 r7 Notice/ s

DEPARTMENT OF ENERGY Effects Abroa f Majoo d r Federal Actionsd an , NEPA. In 1987, DOE Initiated studies, Receip d Financiatan l Settlement Provisions Includin e potentia a th studg f o y l cummula- for Nuclear Research Reactor Fuels tlve environmental effects, to determine the Impact of a 10-year extension of this policy AGENCY: Departmen f Energyto . E programsoDO n . These studie e ongoinar s g and have Identifie a numbed f o Importanr t ACTION: Notice Issues that muse b resolvet d prioo t r extendin e provisionth g f thio s s policr fo y the long term. SUMMARY: The Department of Energy Is amendin e provisionth g currens It f o st policy To provide for continuation of benefi- e providinreceipth d r financiaan fo t g l cial research reactor program o permit d an st settlemen f o U.S.-ot n l spen g rl t research the additional time required for DOE to reactor fuel y extendinb s e dat th y gwhicb e h complet revies a 10-yeaIt e f o w r extensiof no It will receive highly enriched uranium (HEU) this policy, DOE Is amending Its fuel receipt fuels to December 31, 1988. and financial settlement provisions by extendin e effectivth g e datr receipfo e f o t FOR FURTHER INFORMATION CONTACT: U.S.-or Igln HEU research reactor fuels to Loui . WlllettR s , Offic f Nucleaeo r Materials December 31, 1988. To provide for this Production, DP-133.2-GTN, U.S. Department of extension, the following amendment to the Energy, Washington, D.C. 20545, 301/353-3968. Federal Register notice entitled "Receipd an t Financial Settlement Provision r Nucleafo s r SUPPLEMENTARY INFORMATION n NovembeO : , 9 r Research Reactor Fuels," 51 FR 5754, publish- e Departmenth 1982, f Energo t y announcen I d ed February 18, 1986, and as corrected on e Federath l Registe s rextendin wa tha t I t g Marc 7487)R , F 1984 h 1 mades 6,(5 I : until December 31, 1987, Its policy for the receip d financiaan t l settlemen f o U.S.t - 1. Delete paragraph 6.a. and substitute orlgln spent research reactor fuels (47 FR In Its place: 50737). It was determined at that time that there was a continued need in the research r researcFo . h"a reactor fuels described reactor communit a fue r l fo yretur n capa- In 4.a. and 4.c.—December 31, 1988." bilit d e thaan U.Sth yt . Interestn i s limiting worldwide Inventorie U werHE ef o s Troy E. Hade II, served by an extension of the policy. This Acting Assistant Secretar r fo Defensy e extensio s restatedwa n , without changea n I , Programs. February 1986 Federal Register notice that expanded DUE'S fuel receipt and financial [FR Doc. 87-29918 Filed 12-29-87; 8:4] 5am settlement provision o Includt s enrichew lo e d uranium research reactor fuels (51 FR 5754). BILLING CODE 6450-01-M

s determineha E DO d that this noed still exist d thaan s te bes th onc tn I e s agaiI t i n Interese Uniteth f do t State o extent s e th d effectiv e receipeth datr d financiafo an te l settlement for HEU research reactor fuels of U.S. origin. The Department has reviewed the policy extension under the National Environ- s t foun(NEPAha Ac d y mentaan )thaIc l tPo l the extension Itself clearly has no significant Impact. r Exporto subsequen f o s t arrangements Involving nuclear materials are a case-by-cas n o revieweE DO y eb d basin I s accordance wite Guidelineth h r Implementfo s - Ing Executive Order 12114, Environmental

153