Research Reactor Core Conversion from the Use Of

IAEA-TECDOC-324

RESEARCH REACTOR CORE CONVERSION HIGHLF O E YFROUS ENRICHEE MTH D URANIUM TO THE USE OF LOW ENRICHED URANIUM FUELS GUIDEBOOK ADDENDUM: HEAVY WATER MODERATED REACTORS

PREPARED BY A CONSULTANTS' GROUP, COORDINATED AND EDITED BY THE PHYSICS SECTION INTERNATIONAL ATOMIC ENERGY AGENCY

A TECHNICAL DOCUMENT ISSUEE TH Y DB INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1985 RESEARCH REACTOR CORE CONVERSION FROM THE USE OF HIGHLY ENRICHED URANIUM TO THE USE OF LOW ENRICHED URANIUM FUELS GUIDEBOOK ADDENDUM: HEAVY WATER MODERATED REACTORS IAEA, VIENNA, 1985 IAEA-TECDOC-324

Printe IAEe th AustriAn y i d b a January 1985

85-00193 Pleas aware eb Missine th tha l al t g Pages in this document were originally blank pages FOREWORD

e proliferatio th n vieI f o w n concern f highlo e yus s e causeth y b d enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors wil more b l e restricte e futureth n i dGuidebooa , n o k Research Reactor Core Conversio f Highlo e nyUs froEnrichee th m d Uraniuo t m w EnricheLo f o de thUraniuUs e m Fuel IAEA-TECDOC-233( s e s issueth )wa y b d International Atomic Energy Agency in August 1980. IAEA-TECDOC-233 addressed primarily research and test reactors that are moderated by light water. In consideration of the special features of heavy water moderated researc d tesan ht reactors, this documen s beeha t n prepare Addendun a s a d m to IAEA-TECDOC-233 to assist operators and physicists from these reactors in determining whether conversion to the use of low enriched uranium (LEU) fuel design s technicalli s y feasibl r theifo e r specific reactor assiso t d n i t,an making a smooth transition to the use of LEU fuel designs where appropriate.

This Addendum to IAEA-TECDOC-233 has been prepared and coordinated by the International Atomic Energy Agency, with contributions from different organizations e expertTh . s from these organizations have participatea n i d serie f o sConsultants ' Meeting d havan se assiste n preparini d g this text. The safety and licensing aspects of core conversions for both light wate d heavan r y water moderated researc d tesan ht reactor e addressear s d in a separate guidebook currently being prepared under the auspices of the IAEA. CONTRIBUTING ORGANIZATIONS

Argonne National Laboratory ANL United States of America

Atonic Energy Research Establishment HARWELL United Kingdom

Australian Atomic Energy Commission AAEC Australia

Brookhaven National Laboratory BNL United States of America

Chalk River Nuclear Laboratories CRNL Canada

InstituLaun vo Paue- x ltMa Langevin ILL France/Federal Republic of Germany/United Kingdom Japan Atomic Energy Research Institute JAERI Japan

Risp National Laboratory RIS0 Denmark

The contributing organizations would like to thank the following organizations for information referenced in the chapter on fuel development and demonstration: Babcock & Wilcox (USA), CEA (France), CEN-Crenoble (SILOE; France), CEN-Saclay (OSIRIS; France), CERCA (France), CNEA (Argentina) (HFN ,EC R Petten Netherlands)e Th ; , EG&G Idaho (USA) (SAPHIRR ,EI ; Switzerland), GA Technologies (USA), GKSS (FRG-2; FRG), Helsingor Vaerft, now ATLAS A/S (Denmark), HMI (BER-2; FRG), IPEN (IAE-R1; Brazil), KFA-Juelich (FRJ-1, FRJ-2; FRG), KURRI (Japan), NUKEM (FRG), ÖFZS (ASTRA; Austria), ORNL (ORR; USA), Studsvik (R2; Sweden), Texas Instruments (USA) (FRMM ,TO ; FRG), UM (FNR; USA). e IAE gratefu s Th contributionAi e th r lfo s volunteere thesy db e organization thankd san s their experts for preparing the detailed investigations and for evaluating and summarizing the results presente n thii d s Guidebook Addendum. PREFACE GuidebooA Researcn ko h Reactor Core Conversio Highlf o e nyUs Fro e mTh Enrichew EnricheLo f o de d UraniuUs Uraniu e Th mo mT Fuels (IAEA-TECDOC-233) was issued by the International Atomic Energy Agency in August 1980. This document contain widsa e variet informatiof o y physicse th n o n , thermal- hydraulics, fuels fued ,an l cycle economic lighr fo s t water moderated researc d teshan t reactors. n consideratioI speciae th f no l feature heavf so y water moderated research and test reactors (hereafter referre heavs a o t yd water research reactors), this Addendu IAEA-TECDOC-23o mt s beeha 3n prepare assiso t d t operatord san physicists from these reactor determininn si g whether conversion from HEUo t * LEU* fuel designs is technically feasible for their specific reactor, and to assis makinn i t gsmoota U fue LE lf ho design transitioe us e s th wher o nt e appropriate.

The organization of this Addendum follows that of IAEA-TECDOC-233 as closely as possible in order to provide a consistent presentation of the infor- minimizo matiot d repetitioe an n eth f informatioo n n tha commos i t boto t n h heavy water and light water research reactors. Distinctive features of the heavy water reactors are addressed where applicable.

e followinTh g paragraphs provid n outlinea f thieo t si Addendu w ho d an m usee b conjunction i dn ca n with IAEA-TECDOC-233.

. 1 General Considerations

IAEA-TECDOC-233, Section 1.5, provides a summary of the main activities neede preparo t dtypicaa r fo e l conversion. These activitie commoe l sar al o nt research and test reactors. It is possible for the studies that are outlined e performetb o e reactoth y b dr operators/physicists themselves witr o e , th h aid of laboratories which have offered technical assistance.

Section 1 of this Addendum contains descriptions of the design features of heavy water research reactors usin U fue d descriptiongHE lan e th f so reduced enrichment programmes in those countries for which the programmes describeart eno IAEA-TECDOC-233n i d .

2. Reactor Studies and Benchmark Calculations

The core conversion studies presented in this Addendum pertain only to specific reactors, as generic models did not appear necessary or appropriate. The methods and procedures of IAEA-TECDOC-233 have been followed in general, but adaptations have been mad eacr efo h reacto accommodato t r specias eit l design feature operationad an s l requirements.

The results of the specific reactor conversion studies are summarized in Section 2. Detailed information on the methods and procedures used and the results obtainevarioue th r sfo d reactor e presentesar Appendicen i d sA through E. Benchmark problems for both the neutronics and safety-related parameters of heavy water research reactors were defined and calculated in order to compar accurace calculationae th e th f yo l method variouse th use n i ds research centres benchmare Th . k specification (Appendix F-0 idealisen bases a )i n o d d 6 elemen2 t core, surrounde heavy b d y wate graphitd ran e reflectors d operatin,an g at 10 MW. The fuel element consists of four concentric fuel tubes surrounded aluminun ba y m wrapper U tube LE e resultcalculation e Th d .th an f so U HE r fo s fuels are summarized in Section 2.4 and are described in detail in Appendix F. firsa s A t ste corn i p e conversion s recommendei t ,i d that reactor operators/ physicists use their own methods and codes to calculate this benchmark problem, and to compare results.

3. Fuel Development and Demonstration Status

IAEA-TECDOC-233 contains at Section 1.4.2, Chapter 3, and Appendix H, informatio statuse th n no , development potential d commercia,an l availability of fuels with high uranium density as of August 1980. Much of this information is applicabl heavo t e y water reactor fuel ligho wels t a s s tla water reactor fuels. Sectio f thio ns3 Addendum summarize statue th s fuef so l developmend an t demonstratio y 198 Ma r fue 4f fo o l s elemenna t geometries thausee botn ar ti d h light wate heavd ran y water research reactors that currently utiliz U fueleHE .

. 4 IAEA Assistance contactede b IAEe n Th ca A , through official channels provido ,t e coordi- nating assistance between reactor organization d thosan s e laboratories which have offered technical assistance for conversion studies on specific reactors. If necessary, the IAEA can also provide fellowships to visit those laboratories for joint studies on core conversions.

For simplicity followine ,th g definitions have been adopted onlthir yfo s publication. The legal definition of highly enriched uranium is uranium with equal to or greater than 20 wt% 235U.

HEU - Highly Enriched Uranium (>70 wt% 235U)

ME Mediu- U m Enriched 235 % UraniuUwt )5 (4 m Enrichew LELo U- 235d% UraniuUwt ) 0 «2 m CONTENTS Page

1. General Considérations ...... 9 1.1 Introduction ...... 9 1.2 Design Feature Heavf o s y Water Reactors Usin FueU gHE l ...... 9 1.2.1 DIDO-Type Reactors ...... 10 1.2.2 MTR-Type Reactors ...... 12 1.2.3 Other Reactors ...... 4 .1

1.3 Reduced Enrichment Programmes ...... 17 1.3.1 The Reduced Enrichment Programmes of France, the Federal Republic of Germany, Japan, and the United States ...... 17 1.3.Reducee Th 2 d Enrichment Programm Australif eo a ...... 7 .1 1.3.3 The Reduced Enrichment Programme of Canada ...... 18 1.3.Reducee Th 4 d Enrichment Programm Denmarf o e k ...... 9 .1 1.3.Reducee Th 5 d Enrichment Programm Unitee th f deo Kingdo1 2 . m

2. Reactor Studies and Benchmark Calculations ...... 22

2.1 Overview ...... 22 2.2 Results from Reactor Studies ...... 3 2 2.3 Calculated and Estimated MEU and LEU Densities for Conversion of Specific Reactors ...... 24

2.4 Benchmark Calculations ...... 25

. 3 Statu Fuef so l Developmen Demonstration...... d an t 8 .3

3.1 Overview ...... 8 .3

3.2 Status of Plate-Type and Tube-Type Fuel Technology...... 39 3.2.1 U-A1 Alloy Fuel ...... 39

3.2.2 UA1 XUA1-Ad 1an 2 -A1 Fuels ...... 9 3 3.2.3 U3Û8-A1 Fuel ...... 40 3.2.4 U3Si2~Al Fuel ...... 42

3.2.5 U3Si-Al Fuel ...... 3 4 3.2.6 USiAl-Al Fuel ...... 43 3.2.7 UfcFe-Al Fuel ...... 4 4

3.2.8 U02 Caramel Fuel ...... 44 3.3 Statu Pin-Typf so e Fuel Technology ...... 4 4

3.3.1 USiAl-Al and U3Si-Al Fuels ...... 45 3.3.2 U-A1 Alloy Fuel ...... 45

2 Fue3.3.U0 l3 ...... 5 .4

3.3.4 U-ZrHx Fuel ...... 45

References ...... 6 4

APPENDICES

APPENDI XA Enrichment Reduction Calculation DIDOe th .r sfo DR-3 JRR-d .an 2 Reactors (USAL .AN ) ...... 1 5

APPENDIX B Enrichment Reduction Calculations for the DIDO Reactor. HARWELL (UK)...... 107 APPENDIX C Enrichment Reduction Calculations for the HIFAR Reactor. AAEC (Australia) ...... 119

APPENDI D X Enrichment Reduction Calculatione th r sfo JRR-2 Reactor. JAERI (Japan) ...... 145

APPENDI XE Enrichment Reduction Calculatione th r sfo DR-3 Reactor. RIS0 (Denmark)...... 167 APPENDIX F Benchmark Calculations F-0. Specifications ...... 182 (USAL F-lAN )...... 5 .18 F-2. HARWELL (UK)...... 201 F-3. AAEC (Australia) ...... 215 F-4. JAERI (Japan) ...... 231 F-5. RIS0 (Denmark) ...... 239

APPENDIXG Lis Participantf o t Consultantse th n si ' Meeting ...... 251 . GENERA1 L CONSIDERATIONS

1.1 INTRODUCTION

The general concerns discusse IAEA-TECDOC-23n i d proliferatione th n 3o - resistance of fuels and fuel cycles for light water moderated research and test reactors are also applicable to heavy water moderated research and test reactors* Enrichment lesf o se Internationall sar tha% 20 n y recognizea s a d fully adequate isotopic barrie weapono rt s usability certaid ,an n Member States have move minimizo t d internationae th e l trad highly-enrichen i e d uranium. Reduced Enrichment Researc Tesd han t Reactor (RERTR) Programmes have been established to develop the technical means, such as development of w fuel ne desigd san n modifications assiso ,t implementinn i t g reactor conversion U fuelLE o st s whereve r possible with minimum penalties. Related programmes have been establishe numbea n i df othe ro r Member Statess i t I . anticipated that throug continuee th h d effort f thesso e programmes d wit,an h IAEA coordination, many reactors now utilizing fuel element materials and designs less advanced than currently feasible may soon be converted to the U fuel reactorr LE usFo f .o e e s b U fue whosy LE lma f eo conversioe us e th o nt feasible only after significant fuel development, an interim decrease of the enrichmen intermediatn a o t e % (MEUrang45 f eo ) woulworthwhila e b d e improvemen proliferation i t n resistance. Production of plutonium in research reactor fuel may also cause some proliferation concern musd ,considerean e tb wels a duranius la m enrichmenn i t assessin proliferatioe th g n potentia reactorse th f o l fue Al .l elements considere thin i d s report would contain only very small amount plutoniuf o s m upon discharg element)r pe eu P froreactoe g .th m8 ( r

Operators of research and test reactors that use higly-enriched uranium may consider converting their reactors to the use of low-enriched uranium fuels for several closely interrelated reasons. One reason could be the desire to reduce the proliferation ptential of research reactor fuels. A second reason couldesira e b densuro t e e continued fuel availabilite th n i y face of probable restrictions on the supply of uranium enriched to 20% or more in "!>U. A third reason could be the possible reduction in requirement physicar fo s l securit safeguardd an y s measures during fabrication, transportation, storage, and use.

1.2 DESIGN FEATURES OF HEAVY WATER REACTORS USING HEU FUEL

Heavy water research and test reactors using HEU fuel currently utilize numbea f fuero l element designs. These designorganizee b n sca d into three general categories ) concentri(1 : c tube designs characterizes ,a DIDe th O y b d class reactors plate-typR MT ) (2 , e design characterizes sa HFBe th R y b d JRR-e th 2y b reacto d an ) othe Japann A (3 i r US d rreactoe ,an designs th n i r , as characterized by the NRU and NRX reactors in Canada and by the RHF reactor in France. e corfued Th an el element designfive th e f DIDO-claso s s reactore sar summarize Tabln I Tabln i dsimila, e2 . 1 e r design parameter comparee sar d for four reactors currently using fuel elements consisting of curved MTR-type plates. The other designs are summarized in Table 3. The following sections contain brief description f eaco s h reactor type. 1.2.1 DIDO-Type Reactors Six reactor DIDe th O f classo s wer e1955-196e builth n ti 0 perio dthre- e in the United Kingdom and one each in Denmark, the Federal Republic of Germany, and Australia. The one at Dounreay in Scotland (UK) has been shut down, and those at Harwell (UK) and at Juelich (FRG) have had the power increased to 26 respectively, MW an3 2 d , while thos Denmarn ei Australid kan a operate th n i e original power range of 10 MW. The design is similar to the earlier CP-5 reactor in the USA, with a closed cylindrical tank about 2 m diameter x 2.27 m high, containing the 6 fue2 r lo element5 2 cor f rathea o e t sa r givin, opemm n2 ga pitc 15 f o h cylindrical core 610 mm high x 760 mm diameter. The original elements were standard MTR plate type, with 90 g 235u content, soon increased to 115 g U. All reactors have now changed to concentric elements with four fuel tubes 23S, within an outer tube of 103 mm diameter, the maximum port size, and containing an inner diameter m m thimbl fissile 4 5 th f t eo ,ebu content varies, being 150 g for Australia and Denmark, 170 g for the FRG, and 205 g for the UK reactors. Ther e differenceear control-absorbee th n i s safetd ran y equipment now fitted to the reactors. The facilities for irradiation of experiments or for isotope production comprise vertical thimbles inside the fuel elements and in the reflector region. Horizontal facilities are provided for neutron beam work. Whil usee eth reactore s th mad f o e s differ, they have each been optimised by adjustmen f poweo t r leve d fuel-elemenlan t fissile conten givo t e good flux levels for the requirements of their particular combination of neutron beam and irradiation experiments. Fuel elements normally have an average life of 3-6 cycles of about 24 full power days in the reactors. In the higher-power DIDO-class reactors in FRGe th ,d burnablan thK U e e poison usee minimizo sar t d e reactivity-absorption demands on the control absorbers. Fuel loading is adjusted to give the required end-of-cycle reactivity, with sufficient margin to restart in the event of a trip with no complications. The maximum shutdown time before poisoning-out occurs is about 40 minutes, but then it requires about 36 hours before a restart is possible.

10 Table 1. Comparison of DIDO-Type Heavy Water Reactors

Reactor DIDO PLUTO DR-3 HI F AR FRJ-2 UK UK Denmark Australia FRG

Core Description

Power, MW 25. 5 25.5 10 10 23 Core Volume, t 354 360 360 354 348 Ave. Volumetric Power Density, kW/£ 72.0 71.0 27.8 28.2 66.1 Number of Fuel Elements 25 26 26 25 25 Lattice Arrangement Modified Square Square Modified Modified Square Square Square Lattice Pitch, am 152 152 152 152 152

Moderator, Coolant D20 D20 D20 D20 D20 Reflectors D20,C D20,C D20,C D20,C D20,C

Fuel Element Description

U Enrichment, Z 70 70 93 75 or 80 80 and 90 Numbe f Tubes/Elemeno r t 4 4 4 4 4 Fuel Meat Material U-A1 Alloy U-A1 Alloy U-A1 Alloy U-A1 Alloy U-A1 Alloy or UA1X-A1 Clad Material Al Al Al Al Al Fuel Meat Thicknessm ,m 0.72* 0.72* 0.53 0.66 0.66 Clad Thicknessm ,m 0.48* 0.48* 0.47 0.43 0.43 Gap Between Tubes, mm 2.2d 3.0an 99 3.09 and 2.29 3.4 3.38 and 3.29 3.1 and 2.7 235U/Assenibly, g 205 205 150 150 150 and 170 235U Density, g/cm3 0.48 0.48 0.53 0.45 0.42 and 0.47 Ave. Discharge Burnup,Z 45 45 50 54 43

^Average Values 1.2.2 MTR-Type Reactors The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory in the USA is a pressurized MTR-type research reactor that is cooled, moderated, and reflecte heavy b d y fuele s i wate d dran with highly-enriched uraniume Th . HFBR began operation in early 1966 at a power level of 40 MW and was upgraded to 60 MW in September 1982. The reactor close-packea cor s i e d arra 7 element( y s widelement6 * e s long, wit elementh3 s remove eact a d h corner) compose 8 fue2 f lo d elemento tw d san irradiation facilities, each the same size as a fuel element. The active height o in-corcore tw th e e d volumeTh an irra m . abous m £ i e 8 -7 9 t57 core os th fi e diation facilities occupy abou. £ t7 Since 1979, each fuel element has contained 351 g 235U in 18 curved inner plates containing l^Og-Al fuel meat with a uranium density of 1.09 g/cm . The fuel meat has a thickness of 0.58 mm and the 6061 aluminum cladding 3has a thickness of 0.35 mm. Each element also has two unfueled outer plates of 6061 aluminium wit thicknesa h 2.5f so . Previou4mm s designs wit fuele9 h1 d plates and no unfueled plates had 235U loadings of 274 g and 315 g per element with U-A1 alloy fuel (1.0. U) 1U/cm% g wt 0 3;3 The reactor is controlled by 16 neutron-absorbing rods which are located outsid coree e roddividee th eTh ar .s d into setstw o , with eight main rods operating abov core d eighth eean t auxiliary rods operating belo coree th w . The main control rods also serve as safety rods. When the reactor is critical, the two sets of rods are maintained in symmetrical positions above and below the cor minimizo t e e flux perturbation experimentse th n si e shap Th .eacf o e h rod is a hollow right angle with an internal heavy water coolant channel. The principal neutron-absorbing material is Eu2U3 dispersed in a type 304 stainless- steel matrix maie Th .n control rods also contai ndispersioa Dy2Uf no n i 3 the upper portiorode th .f no currene Th day5 td s operatin dayan 5 dow1 p whicn u s ni i W hM g 0 cycl6 t ea 7 fresh element insertee ar s d neacore centee th rth e f aftero r movine th g partially burned elements to the outside. Each element has a residence time of 60 full power days and an average 235U discharge burnup of 46.5%. Sixteen experimental facilitie providee reactore sar th n i d , nine horizontal tube externar fo s l beam experiment d sevesan n vertical thimblet sa central, peripheral and reflector locations for irradiation experiments. The reactor core, reflecto bead an rm tube arrangee ar s enhanco t d low—energe th e y neutron flux in the beams, while decreasing the fast background.

12 Table 2. Comparison of MTR-Type Heavy Water Reactors

Reactor JRR-2 GTRR NBSR HFBR Japan US US US Core Description

Power, MW 10 5 20 60 Core Volumet , 336 209 541 96.8 Ave. Volumetric Power Density, kW/ii 29.8 23.9 37 620 Number of Fuel Elements 24 17 30 28 Lattice Arrangement Hexagonal Triangular Hexagonal Rectangular Lattice Pitcha ,m 134 152 178 -76 Moderator, Coolant D20 D20 D20 D20

Reflectors D20, H20 D20,C D20 D20

Fuel Element Description

U Enrichment,Z 93 93 93 93 93 Element Type MTR Cylindrical MTR MTR MTR Curved Plates Curved Plates Curved Plates Curved Plates Plates/Element 15 Inner 5 Concentric 16 Fueled Uppe7 1 r 18 Fueled 2 Outer Cylinders Unfuele2 d 17 Lower Unfuele2 d p Ga m m 8 17 Sep y b . Fuel Meat Material U-A1 Alloy U-A1 Alloy U-A1 Alloy U308-A1 U308-A1 Clad Material Al Al Al Al Al Fuel Meat Dimensions Thicknessm ,m 0.51 0.51 0.51 0.51 0.58 Width, mm 61.0 48.6-84.4 63.5 62.5 57.2 Lengthm ,m 600 600 597 279 x 2 578 Clad Thicknessa ,m 0.38 Inner 0.38 0.38 0.38 0.35 0.51 Outer Water Chanel Thickness, mm 2.97 3.0 2.69 3.03 2.54 (Ave.) 235U/Elexent, g 195 195 190 Tota0 30 l 351 235U Density, g/cm3 0.67 Inner 0.64 0.61 1.01 1.01 0.21 Outer

Ave. Discharge Burnup,Z 25 25 29.5 55 46.5 1.2.3 Other Reactors

d NR1.2.3.an UX (CanadaNR 1 )

The NRX and NRU reactors at Chalk River, Canada, are large-core, tank-type research reactors usin U fuelgHE . Complete 194 n i 1957d 7an , respectively, both reactors were originally fuelled with natural uraniu lated an m r converted to HEUo otheTw . r NRX-type reactors, CIRU Trombayt Sa , India, buil 1960n i t , R anneadTR r Taipei, Republi Chinaf co , buil 1971n i t , continu fuellee b o t e d with natural uranium. NRX is heavy water moderated, light water cooled, and graphite reflected. It presently operate , usinabout MW a s g5 t2 moderator level contro power lfo r diametern regulationhighm i 2 m ,3. 7 wity 2. e tan9 vertica,b s Th 19 h.i k l through tubes arranged in an hexagonal lattice. The inside diameter of most of thes althoug, mm e w tube 7 tubefe e 5 larger a ar hs i s . NRmoderates Ui cooled dan d with heavy wate combinea n ri d systemd ,an has a light water reflector. It presently operates at 125 MW using sequential control rods for power regulation. NRU has a closed cylindrical tank, 3.5 m hexagonan a s highha m diameten i d 7 l ,an 3. arrangemen y rb 7 lattic22 f o te sites. Permanent tubes, 132 mm inside diameter, extend upwards from the top of the tank through the reactor shielding structure. This arrangement permits on-power refuelling. Fuel assemblie botr fo sh reactor suspendee sar d within individual aluminum flow tubes. Both reactors originally used natural uranium metal fuel. During 1962, NRX was converted to natural UÛ2 fuel, using some highly enriched uranium- aluminum alloy fuel assemblie boosterss a s betweed ,an n 196 197d s graduall8an 0wa y converte U fueHE lo t donly . NRU, meanwhile s converte,wa d directl U fueHE lo t y during 1963-64. The present fuel assemblies for both reactors consist of clusters of 2.7 4lonm g U-A1 alloy (93% fue) 235U ln U i pin s cla aluminumn i d . EacX hNR fuel assembl seves yha n pin stotag wit7 lh54 235U content (0.90 g/cnH) while NRU assemblies have twelve pins with total U content of 491 g (0.63 g/cm ). Equilibrium loadings for NRX require about 23560-70 fuel assemblies, and for NRU3 abou fue0 t9 l assemblies. Typical in-reactor lifetim fueX NR abous li r 5 yearsefo t2. aboud ,an e on t yeaNRUr n botfo rI . h reactor fuee th shuffles li d through several different lattice sites durin lifetimes it g NRUn I . , fuel change generalle sar y made with the reactor operating. In NRX, however, fuel changes must be made with the reactor shut down d moveabl,an e absorbers ("adjuster rods" usee compensat)o ar t d e for reactvity burnup during the three-week operating cycle.

majorite unfuellee Th th f yo d lattice position wida usee r ear s fo d rang e of applications: contro safetd lan y rods, isotope production assemblies, fast neutron rods d hig,an h pressur d temperaturean e loops. Numerous horizontal beam tubes are provided for neutron-beam experiments.

14 1.2.3.2 RHF (France)

The RHF (High F_lux Reactor) is the central facility of the Institut Laue-Langevin TlLL n Grenobli ) e (France). This Institu foundes wa t n 1967i d , originally as a joint German-French project, which the United Kingdom joined s thira d partne n Associate1973e i r Th e Institu.e Scienc th th f d o e s ar an et Engineering Research Council (SERC) for the UK and the Kernforschungszentrum Karlsruhe (KfK r Germany)fo , whil e Frencth e h shar s i dividee d equally between e Commissariath a l'Energit e Atomiqu ee Centr(CEAth d ean ) Nationaa l e d l Recherche Scientifique (CNRS).

The RHF is a very high flux reactor, cooled and moderated by Û20 with a thermal power of 57 megawatts, which has a maximum neutron flux of x ICA 5 1. n/cnrVse e reflectorth n i c . This intens continuoud ean s neutron 5source supplies 35 to 40 different instruments simultaneously, on which research projects can be carried out in the fields of solid state physics, material science, chemistry, biology and nuclear physics.

The thermal neutron flux (peak at 1.2 A in the MaxweHian distribution) is modifie r certaifo d n beam 3 e inclusio t tubeth sourcdm ho y 0 b sa (1 ef no graphite at 2000° K) and a cold source (25 dm3 liquid deuterium at 25° K) givin enchancemenn ga e neutroth f o tn intensit wavelengte th n yi h range respectivelyA 0 4. > X d .an A Neutro 8 0. < n A beam < e availabl4 ar s 0. e from 16 beam port frod neutro0 an sm1 n guides (total lengt meters0 h70 o enablt ) e experimental equipment to be installed which benefits from both a high signal backgrounw lo a flud an xd environment tan° core f s o kD2 Th i eplace .a n i d 2.50 meters in diameter, fixed at the bottom of a H£0 pool, 6 meters in diameter, 14 meters of depth. The pool itself is encased in dense concrete.

The single annular fuel element, containin f uraniuo g k m2 g9. enriche o t d 235 93n i %involut0 Us 28 i mad, f o e e plates (UA1 e fuelth 3 -Ar , fo 1Al-Fe-N i cladding). The reacto s i controller fivy db e safety rod n silver-indium-cadmiui s m alloy cylindricae on , l nicke centrae lth contro n i l d hole burnablro d l an , e poisoning (borone fueth l n platesi ) . The operating day4 cyclshutdow4 y sda s i efollowe2 r 1 fue fo na ly b delemen t replacemen maintenanced an t n additioI . operatin6 o nt g cycle r yearpe s , there is also an annual shutdown of one month to enable necessary maintenance of equipmen e carrieb o t t d out.

Since the first startup at normal power in December 1971, the RHF has completed approximately 60,000 hour f operationo s .

15 Table 3. Comparison of Other Heavy Water Reactors

Reactor RHF NRU NRX France Canada Canada Core Description

Power, MW 57 125 25 Core Volume,I 50 36,400 17,900 Ave. Volumetric Power Density, kH/A 114L140 0 3.4 1.4 Number of Fuel Assemblies 1 ~90 60-70 Lattice Arrangement — Hexagonal Hexagonal 227 Sites 199 Sites Lattice Pitcho ,m - 197 173

Moderator D20 D2Û D20 Coolant D2Û D2Û H20

Reflectors D2<3 0 D2 2H 0, D20, C

Fuel Assembly Description U Enrichment, Z 93 93 93 Assembly Type Annular Pin Pin rolute Plates Cluster Cluster Plates or Pins/Assembly 280 12 7 Fuel Meat Material UAlß-Al U-A1 Alloy U-A1 Alloy Clad Material Al-Fe-Ni Al Al Fuel Meat Dimensions Thicknessm ,m 0.51 5.48* 6.35* Widthm ,m 67 Lengthm ,m 813 2743 2743 Clad Thickness, mm 0.38 0.76 0.76 Water Chanel Thicknessm ,m 1.8 10.5** 11.3** 235U/Assembly,g 8568 491 547 235 U Density, g/cm3 1.11 0.63 0.90 Ave. Discharge Burnup,X 36 85 62

Diameter **MiniauB distance betwee centren npi s

16 1.3 REDUCEJ) ENRICHMENT PROGRAMMES

The IAEA will provide technical assistance to research reactor operators who wish to consider conversion of their reactors from the use of HEU fuel to e contacteU b fuel e IAELE y Th f .Ama o e thd eus throug h official channely b s interested reactor operators to make the necessary arrangements.

Reducee Th 1.3.d1 Enrichment Programme Francef so Federae ,th l Republic of Germany, Unitee Japanth d d ,an State s The Reduced Enrichment Programmes for Research and Test Reactors which are in progress at laboratories in France, the Federal Republic of Germany, Unitee Japanth d d ,an State describee sar Sectio n IAEA-TECDOC-23f i do 3 n1. 3 and are not repeated here. The programmes that are in progress for the heavy water reactors thalocatee tar Australian i d , Canada, Denmark Unitee th d d,an Kingdom are described in this Addendum.

Reducee Th 1.3.d2 Enrichment Programm Australif eo a Intense neutron fluxes for a broad-based research program, and for pro- ductio f radioisotopeno medicar sfo d industrialan l application longa e -ar s standing priority requiremen Australian i t e requireTh . d fluxe providee sar d by the Dido-class reactor, HIFAR, (lUgh Flux Australian jReactor) situated at the Lucas Heights Research Laboratories near Sydney. HIFAR has operated routinely at 10 MW, using highly enriched uranium (> 80% 23SU) fuel for over 20 years, and in this mode has demonstrated a very high level of operational availability and reliability. Australia's geographical isolation from enriched uranium suppliers, fuel fabricators and alternative irradiation facilities comparable to HIFAR, is unique. Reliable and continuing operation of HIFAR is therefore of very great importance, particularly in the medical radioisotope context. Reactor fuel supply line e complesar lond botxn an gi h distanc d timee possibilitan eTh . y that highly enriched uranium might available ceasb source o et a th f s o eea continuing supply of traditional, proven fuel elements for HIFAR, was therefore a cause for grave concern, and led immediately to local studies of means and consequence f reducinso g fuel enrichment. Some aspect f thesso e early studies stile ar generaf lo l interes relevanced tan , particularl thermohydraulice yth s survey calculations summarised at Appendix C2.

Sinc commencemene th e significanf o t t international program enrichmenn o s t reductio researcr nfo h reactors, Australi activels aha y participated, particu- activitiee larlth n i y s coordinate IAEAe n additionth I .y b d , HIFAR-specific studies have continued, and calculations have been updated and refined as further information, especiall relation yi fuelo nt s development s becom,ha e available mose Th .t recent result HIFAf so R studie summarisee sar Appendit a d x Cl. Penaltie operationan i s l flexibilit fued yan l cycle costs wile lb an inevitable accompaniment to any significant reduction of fuel enrichment level in medium/high power research reactors such as HIFAR. The extent of the penalties will be very dependent on the degree of success achieved in advanced fuels development programmes now in progress in several countries. At this stage, development prognoses appear good, but cannot be regarded as assured until results of extensive Irradiation testing programmes on specimen and full size fuel plate varioun i s s element configuration satisfactorile ar s y concluded. Australia intends nevertheless to continue its HIFAR studies, and to keep abreasdevelopmentl al f o t s world-wide, wit viea h reducino t w g HIFAR enrichment,

17 compatible witcommerciae th h l availabilit fulla f yo y proven fuel element, which will guarantee safe and reliable operation at a cost which is acceptable, and not prejudicia HIFAR'o t l s competitive situatios similavi a s r nvi service n i s other type f reactoro s .

Reducee Th 1.3.d 3Enrichmen t Programm Canadf eo a highlf o e yus enrichee Th d uraniu Canadn mi bees aha n mainly limite thao t d t o researc operatioe tw th usee n th i dh f no reactors NRXd an . U ,NR Thes o etw reactors, designed, constructed and operated by Atomic Energy of Canada Limited Chals ait t k River Nuclear Laboratories,, havMW 5 e2 nominad an W M l 5 power12 f so respectively. The annual consumed 235U is some 45 kg and 9 kg, respectively. recognizes Itwa d that, although this uraniu undes mi r strict safeguards contro Canadan i le durinus , s stagel somit al g ef so improvemen proliferan i t - tion risk would (20U occu%LE f 23Sro totan througe Ui us le uranium)th h . With this as an objective, a review of possible fuel materials and fuel rod designs was undertaken in 1978. This review led to the conclusion that due to the high utilizatio f thesno e reactor medicar fo s l isotope productionD ,R& Canadiae supporth r Internationad fo tnan l nuclear programme basid san c science, there was little freedom in altering the design and % content of the exist-

ing fuel rods. Currently, such fuel rods are of U-A 123 alloy clad in aluminum

3 wit maximuha U densit235 m 0.9f o y0 g/cmbasie Th .c characteristicf so the fueeacr lfo h reacto givee rar Tabln ni . e3

Upon analyzing possible alloy and cermet fuels, the US1A1-A1 dispersion fuel was identified as the prime contender such that currenU volumetri235 t c concentrations could be conserved. Due to the well thermalized spectrum in such reactors increasee ,th d 238U fraction acceptabln leada o t s e reduction in reactivity such that onl smalya l increas numbee th fuef n o rei l rodr o s a slightly decreased exit burnup would be required. Calculations indicate that some five extra fuel rods woul requiree b d compensato t d 238 e Uth r penaltyefo , while conserving the exit burnup.

e fueTh l development programme comprises:

developmen fabricatioe th f to USiAl-Af no u*3Si-Ad lan l fuel elements. This includes USiAl and U3$i powder preparation, blending with aluminum powde d extrusioan r n into sillcide dispersion cores whic e thehar n extrusion-clad to manufacture fuel elements. Mini-elements and three full-size NRU fuel rods containing dispersed USiAl have been produced successfully and similar development work with UßSi dispersions is underway. measuremen thermae th f o tl conductivity, chemical compatibility and out-reactor corrosion behaviour of such fuel elements. This programme is in progress. irradiation of mini-elements in NRU to investigate their irradiation behaviour. A power burnup envelope typical of current fuel rods has been defined and the fuel is being tested at the bounding conditions. The behaviour of this fuel will be compared directly with current fuel types and also with a UA1 alloy of higher (1.4 g/cm3) uranium density which would allow the use of 45% EU. fabricatio f threno e full-siz fueU eNR l rods containing USiAl-As lha been complete d theian d r irradiatio begio t 1984n s i nni . Equipment is being installe permio t d manufacture th t U3Si-Af eo l dispersion

.18 fuel elements and the fabrication of three full-size NRÜ rods containing UßSi-Al is scheduled for completion in 1984 with their irradiation to begin in 1985. In parallel with this programme, the performance behaviour of uranium alloy with a uranium density of 1.4 g/cm3 but using MEU is being investigated. The technology for fabrication of such rods is not a significant extrapolation of current methods, and fuel rods of this type have been successfully fabricated. This fue lparticulawoulf U o E usin e % b d g20 r interes NRX-typr fo t e reactorn i s which the experimental load is not as large as that found in the NRX. Current UA1 alloys wit uraniua h m densit 0.9f yo 7 woulg/cmU E usint % bu d3 g20 als o sustai usefuna l loa higt a d h operating flu NRX-typn i x e reactors.

In summary, the USiAl-Al and ^Si-Al fuels are being developed to allow the use of LEU fuels in both the NRX and NRU research reactors. If the current program whicn fue,(i e hth l material manufacturee ar s d using laboratory scale equipment successfuls )i , then production scale equipment will be installed and a 30 rod demonstration irradiation will be performed in NRU. If satisfactory performanc achieves i e d this woul followee b d graduaa y b d l conversion to LEU fuel commencing in the late 1980's. As a backup, and in parallel, the performance behaviour of 45% EU UAl alloy (uranium density 1.4 g/cm3) is being investigated in an irradiation experiment scheduled for completion in 1985.

Reducee Th 1.3.d4 Enrichment Programm Denmarf eo k

1. The Present Situation for Enriched Fuel to Reactor PR 3 Risj& National Laboratory buy % enriches93 d uranium metal whic uses i h d alloy-fuel forUA l element manufactur Helsing<5y b e r Vaerft (HV) (Elsinore Shipyard). It should be noted that the fuel element manufacturing capabili- ties were transferre Danise th o ht d compan yJanuarn ATLAi S SA/ y 1984. heavW M 0 yReacto (1 wate 3 R rrD moderate cooled an d d research reactor), which is the only research reactor in Denmark, burns the HV-fuel elements to about 50Z of the original 23SU content. In the USA, the spent fuel elements e reprocessear credia d an drecoverete giveth r nfo d uranium.

2. Calculations on Low Enriched PR 3 Cores Section 2 of the present report is a summary of conversion calculations on the various heavy water research reactors. According to these calculations expecy ma D 3 Rthermaa t l neutron flux density decreas conversioy eb % 45 o t n enriche dconversioy b fue d 0.9o % enrichet lan 420 o nt d fue 0.8o t l 5 relative flue tth ox condition presene th % enrichen i st93 d core. These figures refer to the experiment holes inside the fuel elements; in the reflector experiment holes the thermal flux density will decrease as 0.97 and 0.95 times the present flux values. The fast neutron flux density will remain nearly unchanged in the core, but it will increase 1% and 2% in the reflector and so will the epithermal neutron flu % enrichmentx20 densitd conversioe an th y % b y45 . o nt Thus, the beam experiments will probably see their thermal/fast flux ratio reduced by factors of 0.96 and 0.93 by conversion to 45% and 20% enrichment, respectively. s desirablIwa t verifo t e calculatione th y w enrichmenlo n so t fuel elements by measurement n sucso h fuel elements.

19 . 3 Test Fuel Element Irradiation3 R D n si

tesx Si t fuel elements were delivered, thre d threean wit e% with45 2 h20 enrichment datA . a lisfollowin e gives th i t n ni g table, wher date th ea also belonging to the 937. enriched fuel elements used until now are presented.

Test Fuel Element Data

In Operation Test Element Test Element

Manufacturer HV HV NÜKEM Enrichment 93% 45% 20% 235U-content 150 g 150 g 180 g 238U-content 11 g 183 g 720 g

Fuel meat Ü-A1 Alloy U-A1 Alloy U308-A1 Cladding thickness 0.47 mm 0.47 mm 0.43 mm Fuel meat U-density 0.57 g U/ cm3 1.18 g U/cm3 2.7 U/cm0g 3

4. Test Fuel Elément Irradiation Schedule

The MEU and LEU test elements (all 1 2345 6 containing flux scanning rigs) were irra- 12 34 diated in core positions Al, B2, and C2. The MEU series was in the core from May 1982 Al to February 1983 U serie, LE followe e sth y b d which lasted until September 1983. The in- B2 stallatio eacf no h elemen s displacetwa e on d operation cycle (four weeks) relative th o t e C2 proceeding element in the same series in order able t b outilizo t e experience th e e gained. The burnup percentages at discharge were 52-58% U tesME te ielementth nU LE e 51-57d th san n %i test elements. 5. Post-Irradiation Examinations Visual examination of separate MEU and LEU fuel tubes in the cutting pond indicated perfect conditions. A more thorough examination in the hot cells would have been performed, in case the visual inspection had unveiled blisters or other irregularities. . 6 Neutron Flux Densit Reactivitd yan y Measurements Measurements of the fast and thermal neutron fluxes in each of the test elements were mad variout ea s e irradiationsstageth f so measurementd an , s reactivite oth f y changes wer withdrawae insertioe th th madt a t e a d f lo nan each test element e datTh .a from these measurements were evaluated an d presented at the RERTR-program meeting in Tokai-Mura, Japan, 24-27 October, 1983. The results confirmed the calculations in appendices A, E, and F-5, summarized in Section 1.3.4.2. The detailed results will be presented in the IAEA guidebook on the safety and licensing issues related to research reactor core conversions (to be published). . Feasibilit7 y Stud Potentian o y FueU 3 lLE lR P Typr fo e The consequences to reactor performance of a conversion to LEU fuel operatioe fo wer3 th r R eD revealef no calculatione th y b d measurementd san s described above. Furthermore, to evaluate the consequences for the reactor safety, the manufacturing process and the fuel cycle costs, a feasibilty

20 study was started late 1983. The objective was to establish a basis for the decision as to which fuel type would be the best suited for a potential LEUo conversio t l aspect ,3 al R D sf no take considerationn ni ; to_ evaluate e fabricatioth n technolog d costsan y d toan ,^ prepare detailed planninf o g subsequent research programme for the period 1984-1988. In this context, the oxide fuel type was omitted due to the safety implications of a LOCA on this fuel type in a reactor like DR 3. Consequently studiee ,th concentratee sar suicidn do e fuel wite hth same geometr presene UAl-alloth n o U fue s a yd tHE lan y fuel with changed geometry: thicker fuel meat and possibly three fuel tubes instead of four. Decision on the preferred fuel type is expected in late 198A. e feasibilitTh y stud beins i y g performe cooperation i d n wite th h Danish fuel element produceRERTR-groue th d ran ANLt pa .

Reducee Th 1.3.5d Enrichment Programm Unitee th f deo Kingdom Th K produceeU Harwele sth fuer llfo reactors, DIDs PLUTOd ha Oan d ,an supplie othee th rdl DIDal fue o Ot l class reactor numbea o t wels f a so rs la light water reactors. reprocesses i fueR K recoveree U MT th e l l d th use Al an dn i dd enriched uranium recycle blendiny b d g wit % enricheh93 d e uraniumth o t d , le whic s hha mor, or e % recently80 usf eo % enriche,70 fueR d MT fuel ll productioAl . s nha used the uranium-aluminium metal alloy route, with pure aluminium cladding. To make available replacement varioue th r s fo s R fuel typeMT f ,o s using reduced enrichment, it was realised that the design and thermal-hydraulic performance had in every case received long development and proving, and s thereforwa tha t i t e necessar e higheus o t yr uranium densitsame th e n i y geometrical design of fuel. An assessment indicated that while the metal alloy process could be extended sufficiently to allow use of MEU (45% enrichment), a new technology U (20LE %f o enrichment)e necessarus s e wa th r yfo . Taking accoun availf to - able facilitie d reportesan d operating experience frohige th mh duty reactors, the U3Û8-A1 cermet clad in aluminium alloy was selected as most suitable. The first phase of work has produced MEU elements of both metal alloy and cermet type and these have been irradiated in the PLUTO reactor at Harwell which subjects them to the full burnup of 50% or more in three or four cycles each of 24 days. The next phase of work is producing LEU elements by the cermet technology, and it is expected that these will be available for proving irradiations during 1984. Before any reactors are converted to a full core of a new type, it will be necessar re-examino t y reactoe th e r safet e usereffece yth th cassn d o t ean of the reactor. These considerations may affect the specification of the new type fuel or the economics of operation, so that the initial objective of a replacement of virtually identical design may not be adequate and further developmen reactoe th f r fue o to ry therefor lma e requiredb e .

21 . REACTO2 R STUDIE BENCHMARD SAN K CALCULATIONS

1 OVERVIE2. W

Independent calculations have been performed by a number of research centrefeasibilite th n o s r conversiofo y f specifino c researc d teshan t reactors from HEU fuel to both MEU and LEU fuels. The methods and procedures outline IAEA-TECDOC-23n i d 3 have been followe generaln i d adaptationt ,bu s have been madeacr fo eh reacto accommodato t r specias it e l design featured an s operational requirements. The specific reactors include the DIDO reactor in the United Kingdom, the HIFAR reactor in Australia, the DR-3 reactor in Denmark, and the JRR-2 reacto Japann ri . Detailed description calculationae th f so l methode th d san results of these studies are provided in Appendices A through E. Most of the calculations with MEU and LEU fuel were performed with fuel element geometries identical with those in current use. In these cases, the uranium density in the fuel meat was increased until the reactor had the same lifetime and end of cycle excess reactivity as the current HEU core. In some cases where the clad is thicker than the 0.38 mm used in many other light wate heavd ran y water (see Tabl reactors) e2 , calculations were performed with 0.38 mm clad and increased fuel meat thickness such that the overall thickness of each fuel tube was preserved. In these cases the required uranium density was reduced in approximate proportion to the increase in fuel meat volume. Since the thickness of the fuel tubes was not altered, the steady-state thermal-hydraulic safety margins with MEU and LEU fuels are expected to be very similar to those for the current HEU cores. somr Fo eU fue LE possiblreactorse lb e elementy us ma o t et , i s with thicker fuel meat and narrower coolant channels without significant changes in the steady-state thermal-hydraulic safety margins. Survey-type thermal- hydraulics calculation reactoro tw r presentee sfo sar Appendicen i d. D d an sC benchmarf o t Ase k calculation idealizen a r sfo d configuratio s alsnha o been performe fivy b d e laboratorie r botfo sh neutronic d safety-relatesan d parameters e purposTh . f theseo e calculation checo t w closel s kho i s e th y results obtained by the various laboratories compare when the calculations are identicar rufo n l conditions e reactoTh . r specifications use thesn i d e calculation realistice meant b no o result e t e th ar s d ,san usee shoulb dt dno to draw conclusions about the performance of specific reactors. Comparison of the results gives an indication, however, of the reliability of the methods and of possible biases. The results of the benchmark calculations are summarized in Section 2.4. Detailed description calculatione th f o s providee sar n i d Appendix F. The choices of optimal conversion strategy, fuel element geometry and 235U loadin specifir fo g c reactor conversions will depen individuan o d l assessments of the trade-offs among economic, performance, safety and licensing issues. For example fuele ,th s thaavailable tar conversionr efo particulaa t sa r time will statue depenth f thein o so d r development, demonstratio d licensabilitynan . Economic consideration choosinn i s fuega l element geometr uraniud an y m density for light water reactor discussee ar s Appendin i d IAEA-TECDOC-233f o xI . Fuel cycle costs are not addressed in this Addendum, but the methods presented in IAEA-TECDOC-23 startina usee b s a dn 3ca g poin d modifiean t updated an d s a d appropriat specifir fo e c cases. If the type of fuel is changed, the performance of the new fuel will need to be examined closely. For example, thermal capacities and thermal conductivi- ties are different for different fuel compositions and weights.

22 e safet Th d licensinyan g aspect corf so e conversion botr sfo h light water and heavy water reactor addressee sar separata n i d e guidebook currently being prepared unde auspicee IAEAe th r th .f so

2 RESULT2. S FROM REACTOR STUDIES e overalTh l conclusions drawn frostudiee th m thin i s s Addendu heavn o m y water reactors are In very good agreement with each other and are very similar to those for the light water reactors studied in IAEA-TECDOC-233. The excep- tio thas increase ni th t 235n ei U loading require maintaio t d same th ne cycle f cyclo d elengten exces d han s reactivit smalleheavs e i yth yr rfo wate r reactors becaus theif eo r open-pitch design becausd san e heavy watemora s eri efficient reflector than either light water or graphite. The key results are summarized in Tabl. e4 Tabl . eCalculate4 d Performance Results MEU/HEU LEU/HEU U Loading Rati Fresr ope h Element 1.02-1.05 1.07-1.12 23S Average Flux Ratios in Core Fast O5.53 keV) 0.99-1.01 0.99-1.01 Epithermal 0.99 0.97-0.99 Thermal «0.62 ) 5eV 0.92-0.95 0.83-0.88 Thermal Flux Rati Peat a oHeavn I k y Water Reflector 0.95-0.97 0.91-0.95

Since the calculations for each reactor were done using its operational cycle length with HEU, MEU, and LEU fuels, the number of elements consumed peaverage rth yead ean r discharge burnup (measure MWdn i d ) wersame th ee with threl fuelal ef so enrichments . However percentage ,th e 235U discharge burnups in the MEU and LEU elements were lower because the 235U contents of the fuel elements were largeenerge th somd e ran f yth o e productioo t e du s nwa burnu f plutoniumpo . Concern was expressed by reactor operators that the flux reductions with LEU, particularl in-core th f yo e thermal flux, could adversely affece th t utilizatio e reactorth f n o d tha e fissil,an th t e conten fuelf o t , operating power level and operating cycle might have to be re-optimized within the limitations set by national authorities. Economics would probably also be affected. These matters will be individual to each reactor and are considered outside tb o scope th ef thieo s Guidebook Addendum.

23 2.3 CALCULATED AND ESTIMATED MEU AND LEU DENSITIES FOR CONVERSION OF SPECIFIC REACTORS The minimum uranium densities with MEU and LEU fuels that were computed for conversio specifie th f no c reactors include thin i d s Addendu showe ar m n inr simplicity TablFo . s beee5 ha nt ,i assume d tha geometro n t y changes are made in the current HEU fuel elements, even though geometry changes are feasibl severan ei l cases. Onlenrichmene th y densitd uraniue tan th f o ym in the fuel meat have been altered for the cases tabulated here. With these loadings calculatione th , s indicate thae reactorth t s would have approximately the same cycle length, experimenta f cyclo l d eloadsen exces d ,an s reactivity as the current HEU cores. Explicit calculation t includeno e thin sar i d s Addendue th somr f fo meo reactors liste Tablen i d s 1-3. Rough estimate minimue th f o sm uranium densities required for conversion of these reactors are also included in Table 5 % increas4 e basea U th n o dn loadinei increas% g10 wit a fueU d ehME l an 235with LEU fuel. Detailed calculations similar to those described in this Addendum would have to be performed in order to obtain more accurate estimates. However listee ,th d uranium densities coul usefue b dguidelina s la n i e beginning these calculations.

Table 5. Calculated and Estimated Minimum Uranium Densities Required for Conversion of the Heavy Water Reactors U FuelLE d s an U ListeME o Tablen t i d 3 s1- Current Minimum Uranium Uranium Current Density, g/cm3 Density, Enrichment Reactor Country g/cm3 w/o 235U MEU (45Z(20ZU LE )) Calculated DIDO UK 0.68 70 1.1 2.7 HIFAR Australia 0.56 80 1.0 2.3 DR-3 Denmark 0.57 93 1.2 2.8-2 .9 JRR-2 Japan 0.69 93 1.4 3.4-3 .5 Estimated PLUTO UK 0.68 70 -1.1 -2.7 FRJ-2 FRG 0.52 90 -1.1 -2.6 GTRR USA 0.66 93 -1.4 -3.4 NBSR USA 1.09 93 -2.3 -5.6 HFBR USA 1.09 93 -2.3 -5.6 RHF France 1.20 93 -2.6 -6.1 NRU* Canada 0.68 93 -1.5 -3.5 NRX* Canada 0.97 93 -2.1 -5.0

*These reactors also allow more fuel elements to be loaded as an alternative to increasin 235e Uth g conten elementr pe t . Channel power limitations could preclud increasn ea 235n ei U conten elementr pe t .

24 2.4 BENCHMARK CALCULATIONS Introduction

As a validation exercise in support of the calculations being carried out by participant Consultantse th f o s ' Meeting enrichmene th n o s t reduction proposals for their individual reactors, a series of benchmark calculations has been devised. These are based on an idealised model of a PLUTO type heavy water research reactor n intercomparisoA . presentes ni dresulte th her f eo s benchmare oth f k calculations received froparticipantse th m e calculationTh . s requested are aimed at providing a thorough test of core physics computing code datd severan san i a l area f relevanco s enrichmeno t e t reduction. These include the accompanying changes in flux levels, influence on reactivity balance, and safety related effects Including temperature and void coefficients. The specifications of the benchmark reactor configuration and operating conditions and the various parameters to be computed appear in Appendix F-0. Intercomparison of Results The first task in the benchmark calculation is the computation of the infinite lattice cell multiplication as a function of the mass of 235U burned for the two different enrichments and fuel element fissile contents - 93% enrichment with 150 g 235U and 20% enrichment with 167.5 g 23SU. The results plotte Fign i d . sho1 wvera y good measur f agreemeneo t particularl% 93 r fo y enrichment and basically give confidence that the burnup codes used to compute the buildup of fission products are consistent and that the 235U data are similarly reactive. The fact that the curves for low enrichment do not show quite such good agreement indicates that there are differences among participants in their treatment of 238U absorption, which is central to the question of enrichment reduction t shoulI . notee b d d s plottethai Fign » i tk .d, 1 against mass of 235U burned which includes that removed by (n,y) reaction as well as that fissioned. It should be further noted that because of the conversio 238 f 239o no Ut Pu , which then undergoes fission mase 235f ,th o s U burned is not in proportion to the integrated power. Tables 6 and 7 show the variation in concentration of selected isotopes enrichment% 20 durind an fuee th g% l , 93 lif respectivelyr efo . Excellent agreement is observed In the case of the 135Xe, but the 239Pu and total Pu concentrations show approximatel scatte% 10 ± y r abou meane th t . Broadly these are consistent witpercentage figuree th th h r fo s 238 f o eU convertede Th . results in Tables 6 and 7 are plotted (with the exception of 238U) in Figs. 2 shoo measure t an wth 3 d f agreementeo . The next stage after the infinite lattice cell calculations is an X-Y calculatio reactoe th f no r witfuee th hl elements having reache U 3 depletio" d n O Q C conditions as detailed in Appendix F-0. Beginning and end of cycle (BOG and EOC) reactivity value showe sar n diagraramaticall Fign botr i y fo h. 4 enrichments. The overall agreement is good, with the exception of the Harwell results which are less reactive than the rest by 2-3%. However, since this disagreement applies equally to both high and low enrichment cases, one can assume that the Harwell conclusions concerning chang enrichmenf eo t wil unaffectede lb . Figur givee5 s the flux distributions along the x-axis at the core mid-plane in three energy groups for 93% enrichment. Apart from some small differences in the size of the reflector peak in the thermal group, the results are in excellent agreement. Figur showe6 changee th s reacton i s r performanc flue n termth i ex f o s ratio between 20% and 93% enrichment. The participants predict an almost Identica e thermal th curv r lfo e flux rati t ther some bu o ar ee disparitien i s the reflector, particularl epithermae th n i y l group. The safety-related part of the benchmark was based entirely on fuel elements wit uniforha sake simplicityf th eo m 235 g r burnuU0 fo 4 r f po Fo . both 93% and 20% enrichments, lattice cell calculations were performed in

25 which the whole cell was at 20°C (the base case), followed by cell calculation whicn i s e fue d coolanth han l t includin centrage th tha n i tl facility wer 50°Ct a e hypotheticaA . l configuratio whicn i ncoolane th h t remainet a d 20°C whil fuee rised th e lha 300°o t ns als Cwa o computed, wit furtheha r case being considere withiÛ whicD2 n i de fuee nth th h l element coolant channels onlreduces ywa 235% densitn 20 U i dd an 20%y b y% . 93 sho9 Table r d wfo an s8 enrichment, respectively, the infinite lattice cell reactivities, followed by keff values for whole reactor X-Y calculations. It should be noted that the coolant density reduction was only applied to the central two fuel elements. e reactivitTh y changes frobase th me cas alse ear o given, together wite th h differential values with respec temperaturo t d percenean t voidage change. Agreement among the participants is excellent throughout, the accord being very pleasin respecn Dopplee i g th f o tr coefficiene Uth f whicto s i h largely responsible for 238the reactivity change on raising the fuel to high temperature. optionae th n I lbenchmare partth f o s k exercise, three participants carried out a comparison of reaction rate ratios between 20% and 93% enriched cores. Ratios at core centre and at X m 532 mm in the X-Y model for uniform 40 g U burnup fuel elements are reported in Table 10. The agreement for all isotope 235 s computed is excellent.

Conclusions The results of the benchmark calculations performed by five participating centres have been compared in Tables 5-9 and Figs. 1-6. They show in general an excellent measur f agreemeneo t w exceptionwitfe ha attributee sb than ca t d to minor differences in the calculations and data. This study has been of great valu givinn i e g confidenc participante th o et s thacalculatione th t s they have carried out on their own reactor systems are realistic and that the implications of enrichment reduction can be adequately predicted.

26 FIG. I VARIATION IN INFINITE LATTICE CELL CALCULAION UlTH U235 BURN UP

1.9 —————. A.A.E.C __._. J.A.E.R. I _ _ _ R. I. S. 0 ———._ A/M. L ————— HARWELL

1.8

K-INF

10 20 30 50 U23G 5 BURNP TU

27 Table 6 93% Enrichment 150g U-235 per element

Number densities of important nuclides (Nucler ipe 10-24 cm:') g U-235 burned Nuclide AAEC JAERI RISC) ANL HARWELL 0 Xe-13. 0 5 0 0.0 0.0 0.0 0.0 20 1.291E-08 1. 269E-08 1.308E-08 1. 299E-08 1.283E-08 30 1.197E-08 1.171E-08 1.219E-08 1.192E-08 1.191E-08 40 1.102E-08 1.082E-08 1.128E-08 1.104S-08 1.098E-OS 45 1.054E-08 1.029E-08 1.037E-08 1.067E-08 1.051E-08 60 9.103E-08 9.077E-09 9.452E-09 9.404E-09 9.095E-09

0 Pu-239 0.0 0.0 0.0 0.0 0.0 20 6.710E-07 5.220E-07 4.407E-07 5.875E-07 5.878E-07 30 9.335E-07 7.320E-07 6„796E-07 8.147E-07 8.162E-07 40 1.149E-06 9. 044E-07 8.534E-07 9.986E-07 1.003E-06 45 1.236E-06 9.746E-07 9.933E-07 1.078E-06 1.081E-06 60 1.450E-06 1.148E-06 1. 100E-06 1.260E-06 1.255E-06

0 Total Pu 0.0 0.0 0.0 0.0 0.0 20 7.096E-07 5.500E-07 4.938E-07 6. 200E-07 6.202E-07 30 1.018E-06 7.952E-07 7.341E-07 8.865E-07 8.879E-07 40 1.296E-06 1.016E-06 9.521E-07 1.123E-06 1.128E-06 45 1.421E-06 1.114E-06 1.1478E-06 1.234E-06 1.237E-06 60 1.763E-06 1.389E-06 1.321E-06 1.526E-06 1. 522E-06

0 U-238 9.679E-05 9.679E-05 9.679E-05 9.679E-05 9.679E-05 20 9.600E-05 9.618E-05 9.619E-05 9•611E-05 9.593E-05 30 9.559E-05 9.586E-05 9.588E-05 9.575E-05 9.550E-05 40 9.517E-05 9.553E-05 9.556E-05 9.540E-05 9.507E-05 45 9.496E-05 9.537E-05 9.523E-05 9.521E-05 9.485E-05 60 9.431E-05 9.484E-05 9.489E-05 9. 465E-05 9.418E-05 percent los U-23n si 8 whilU-23g 0 burne6 s e5i d 2.6 2.0 2.0 2.2 2.7

28 Table 7

20% Enrichment 167.5g U-235 per element of important nuclides (Nucle 10~^r ) ipe m ^c g U-235 burned Nuclide AAEC JAERI RIS0 ANL HARWELL

0 Xe-135 0.0 0.9 0.0 0.0 0.0 20 1.485E-08 1.465E-08 1.483E-08 1.479E-08 1.470E-03 30 1.404E-08 1.379E-08 1.407E-07 1.385E-08 1.390E-08 40 1.321E-08 1.297E-08 1.329E-08 1.307E-08 1 . 308E-03 45 1.279E-08 1.248E-08 1.250E-08 1.273E-08 1.266E-08 60 1.149E-08 1.136E-08 1.170E-08 1.157E-08 1.137E-08

0 Pu-239 0.0 0.0 0.0 0.0 0.0 20 «i 1.369E-05 1.051E-05 1.203E-05 1.189E-05 1.2401E-05 30 1.935E-05 1.499E-05 1.751E-05 1.680E-05 1.7945E-05 40 «• 2.423E-05 1. 887E-05 2.223E-05 2.102E-05 2.3076E-05 45 *• 2.631E-05 2.057E-05 2.624E-05 2.291E-05 2.5488E-05 60 3.175E-05 2. 501E-05 2.956E-05 2.761E-05 3.2208E-05

0 Total Pu 0.0 0.0 0.0 0.0 0.0 20 it 1.4408E-05 1.1024E-05 1.258E-05 l.249E-05 1.2848E-05

30 t* 2.0933E-05 1.6155E-05 1.880E-05 1.814E-05 1.8667E-05 40 i* 2.6997E-05 2.0934E-05 2.455E-05 2.336E-05 2.4096E-05 45 11 2.9814E-05 2.3171E-05 2.985E-05 2.585E-05 2.6661E-05 60 *• 3.7768E-05 2.9565E-05 3.471E-05 3.270E-05 3.3743E-05

0 U-238 5.744E-03 5.744E-03 5.744E-03 5.744E-03 5.744E-03 20 it 5.728E-03 5.7323E-03 5.729E-03 5.730E-03 5.728E-03 30 *• 5.719E-03 5.725E-03 5.721E-03 5.723E-03 5.720E-03 40 •• 5.711E-03 5.718E-03 5.713E-03 5.716E-03 5.713E-03 45 *• 5.706E-03 5.715E-03 5.704E-03 5.711E-03 5.709E-03 60 n 5.692E-03 5.704E-03 5.696E-03 5.702E-03 5.696E-03 percent los Ü-23n si 8 whilU-23g 0 e6 5 burned 0.91 0.70 0.84 0.730.84

29 FIG .VARIATIO2 135-XEN I N , 239-P TOTAD UAN CONCENTRATIOU LP N WITH 23S-U BURN-UI ENRICHMENT9Ï T PA S

1.6

X 1

d C E. A. A I —-• J.A.E.R.I — - R. I.S.tf —— A.N. L HARUELL

0 2 0 0 6 0 4 0 2 40 20 40 6C

G 235-U BURNP U T

30 FIG.3 VARIATIO 135-XEN I N , 239-P TOTAO UAN L PU-CONCENTRATION WITH 235-U BURN-U 1 ENRICHMEN20 T PA T

1.6

1.4

1.2 o

X V-0 I

0.8

C . E A. .A J. A. E. R. I. g 0.6 R. I. S. 0. u A.N. L HARWELL

0.4

0.2

0.0 20 40 60 0 20 0 6 " 40 20 40 6C G 235-U BURNT UP

31 FIG.4 BENCHMARK REACTOR OPERATING CYCLE REACTIVITIES

t.; S.Ü

93Z EK:!CH'OT

it J •

cl -c 20Ï ENRICHMENT . u3 . 1 - — 2 s t. IS -*\*\ e - g » J r « 1 < V ' 1 < 5 ' S I j * ^ ^ 1 i " 1 i - -! a: Î L T " * " d 3 a y Ü -si > <- ï d 1 I II l x UJ CO s s •< Œ

5» fa KEY u K BEG NI INCYCLF GO E REACTIVIT« Y (BOO a END OF CYCLE REACTIVITY (EOC) KEY .&p (BÜC-EOC) 931 ENRICHMENT + 20Z ENRICHMENT

1 C

32 FIG. 5 RADIAL FLUX DISTRIBUTION AT CORE HIDPLANE FOR BEGIIN1NG OF CYCLE,931 ENRICHMENT BENCHMARK REACTOR AT 10 MW

GRAPHITE

—————— A.A.E.C — R.I.S.J — — — ) .—————— A. N. L —————— HARWELL

GROUP 3 (0 - 0.625 EVI

0 12 0 11 0 10 0 9 0 8 0 7 0 6 0 5 O A 0 3 0 2 0 1 0 RADIAL DISTANCE FROM CORE CENTRE (CM)

33 FIG. 6 RADIAL DISTRIBUTIO FLUF O N X RATI (200 O) /f (93) BETWEEN BENCHMARK REACTOR ENRICHE1 93 D SAN D UIT 1 FUEH20 L 1.05 1.00 r Ratio of Then«. I Flux«* <0. 625 oV rati f faso t fluxe3 Ica5 V. s>5 cor» 020 graphita

020 graph]t*

• S

0.95 1.00 0 12 0 10 0 8 0 6 0 4 0 2 ro 1.05 c> core 0 D2 graphe tc -0- A.A.E.C o ___ R. I.S.j» CNJ _.——— A.N.L HARWELL o "0.90 I—I r— 1.00 -<

thermaI Ratiep f o l fluxesO2 .6 0.625.5O T V V 85a ke

0.95 •-0.85 0 0 12 0 10 0 8 0 6 0 4 0 2 20 40 60 80 too 1: RADIAL DISTANCE FROM CORE CENTRE,CM

34 Table 8

93% Enrichment Reactivity Coefficient Results

AAEC JAERI L AN RIS0 HARWELL

A Infinite Lattice Calcs 1. kgo base case at 20° 1.70057 1.68810 1.69192 1.69617 coolan° 50 2, .d tko« an 1.69870 1.68608 1.69026 1.69430 fuel 3. koo 300°fuel 1.70039 1.68774 1.69178 1.69600 4. kgo 20% coolant void 1.69909 1.68591 1.69005 1.69446

B X-Y Calcs 1. k-eff base case at 20°C 1.14962 1.1479 1.15413 1.12947 2. k-eff 500C coolant 1.14433 1.14897 1.12399 and fuel

3. k-eff 300°C fuel 1.14956 1.15410 1.12924 4. k-ef coolan% f20 t void 1.14612 1.15086 1.12598 (central 2 elements)

C A/ »fro% m Base Case 2. 500 coolan fuet& l -0.403 -0.401 -0.488* -0.389 -0.431

3. 30QOC fuel -0.005 -0.001 -0.045* -0.002 -0.018 4.coolan% 20 t void -0.266 -0.287 -0.315* -0.246 -0.274 (central 2 elements)

D. Mean Slope A ft %£k/k per °C or % void x 10~6 2. 50°C coolant & fuel -134 -134 -163 -130-144 3. 300°C fuel -0.2 -0.03 -1.6 -0.07-0.6 4. 20% coolant void -133 -143 -157 -123-137 (centra elementsl2 )

^Results obtained directly from base case by perturbation theory

35 Table 9 20% Enrichment Reactivity Coefficient Results

AAEC JAERI RIS0 ANL HARWELL A Infinite Lattice Gales

1. koo base case at 1.61652 1.59660 1.60043 1.62012 20°C

2. koo 500 coolant & 1.61365 1.59363 1.60552 1.61724 fuel 3. '<<„ 30QOC fuel 1.60730 1.58708 1.60027 1.61096

4. 20% coolant void 1.61370 1.59268 1.60513 1.61748

B X-Y Gales K-ef. 1 f base cast ea 1.12903 1.1266 1.13393 1.11585 20° C

2. K-eff 50°C coolant 1.12398 1.12860 1.09211

3. k-eff 3000 fuel 1.12391 1.12862 1.11026 k-ef. 4 coolan% f20 t void 1.12579 1.13066 1.11256

G A P % from Base Case

2. 50° coolant & fuel -0.398 -0.398 •0.457* -0.417 -0.424

3. 300°C fuel -0.403 -0.381 •0.488* -0.415 -0.451

4. 20% coolant void -0.255 -0.275 •0.296* -0.255 -0.256 D Mean Slope AP %<îk/k per °C or % void x 10'6 coolan0 2.50 fue& t l -133 -133 •152 -139 -141

3. 3000 fuel •16.0 -14.9 -16.1 coolan% 20 . 4 t void -127 -137 •148 -128' -128 (centra elementsl2 )

*Results obtained directly from base case by perturbation theory

36 Table 10 Absorption Reaction Rate Ratios Z (20%)/i|> E (93%)

Edited Over Central Experiment Space

Core Centre 532 mm from Core Centre Isotope

ANL JAERI RIS0 ANL JAERI RIS0

U-235 0.84 0.85 0.85 0.94 0.94 0.92 Mo-98 0.98 0.98 0.99 0.996 B-10 0.84 0.84 0.84 0.94 0.94 0.92 Co 0.86 0.87 0.94 0.94 Si 0.84 0.94

N-14 0.84 0.85 0.94 0.94

37 3. STATUS OF FUEL DEVELOPMENT AND DEMONSTRATION

3.1 OVERVIEW The status of research reactor fuel development as of March 1980 was discusse Section i d n 1.4.2, Chapte Appendid an IAEA-TECDOC-233f , ro 3 xH . Much of this information is applicable to heavy water reactor fuels as well as to light water reacto heave th r y f fuelso wate l Al r. reactors listen i d Tables 1, 2, and 3 utilize plate-type designs except for NRU and NRX which utilize a pin-type geometry. The plate-type designs comprise both rectangular MTR-type elements (or an annular element) and DIDO-type concentric tubular elements. Fuel tubes for the latter design are manufactured either by electron beam welding from three pre-clad curved plateco-extrusioy b r so o nt produce clad fuel tubes in a single operation. Since 1980, ther bees eha n significant progres developmene th n si d an t demonstratio fuelf no s with high uranium densities. This section contains informatio statu e y 198th Ma fuef n 4sf o o o ls na developmen demonstrad an t - tion for both light water and heavy water research and test reactors that currently utilize HEU fuel. It is important to note that individual specifications (e.g. highly curved plates fabricatior )o n methods (e.g. co-extrusion) for heavy water reactors require development and testing to prove that the element manufacturee b n ca s d reliabl productioa n yo n basis.

Summary - Status of LEU Fuels - May 1984

Probability Approximate Date Fuel-Type g/cm3 for Success of Licensability U-A1 Alloy 1.1,1.2 Assured 1981,1983

UA1X-A1 2.2 Assured 1983

U30s-Al 2.7-3.2 Extremely High 1984-1985

3.7 Extremely High 1985 4.8 Very High 1985 4.8-5.5 High 1986-1988

5.5-7.0 High 1986-1988

UÛ2 Caramel 9.1 Assured 1980

USiAl-Al* 3.2-4.8 High 1986-1988

U3Si-Al* 3.2-4.8 High 1986-1988 U-ZrHv 2.2 Assured 1983 3.7 Extremely High 1985

*Pin-Type Geometry

38 3.2 STATUS OF PLATE-TYPE AND TUBE-TYPE FUEL TECHNOLOGY

3.2.1 U-A1 Alloy Fuel Full-Sized Elements

A full-sized DIDO-type concentric-tube element fabricated by the UKAEA wit U-AU hME 1 alloU/cmg s irradiated1 ywa 31. fued lan 1»2 during e 198th 1n i PLUTO reacto Harwelt ra averagn a o lt e U burnu 52%f po elemene .Th s twa fabricate normae th o lt d production 235 specification usin traditionae th g l route of rolling plate d producinan s g fuel tube electroy b s n beam weldin f preformo g - ed plates for assembly into concentric-tube fuel elements. Post-irradiation examination (PIE f thi)o s element showed excellenn i s wa tha t ti t condition. 2 During 1980 and 1981, the UKAEA also demonstrated that it is possible to extend the plate production process to routinely fabricate alloy fuel with a uranium density up to 1.34 g/cm3 (35 w/o U) without further development. Three full-sized concentric-tube elements wit U-AU hME 1 alloy fued lan U/cmg 2 31. fabricate Helsingfiy b d r Vaerft usin plate th g e route were successfully irradiated3 in the DR-3 reactor at the Ris«i National Laboratory to average 235% betweeU58 - 198burnupy 2 nMa Februard 5 2 an f o s y 1983. Visuaf individuao E lPI l fuel tubes foun excellenn i d e b the o t mt condition.

3.2.2 UA1 XUA1-Ad 1an 2 -A1 Fuels

Small Test Plates

PIE have been complete miniplate2 2 n o d s with UA1X-A1 fuel (1.5- 2.3 g U/cm with MEU and 1.9 - 2.5 g U/cm with LEU) that were manufactured by E6&G Idaho3 , NUKEM, and the CNEA. All o3f these plates were irradiated1* in the Oak Ridge Research Reactor (ORR) at ORNL to measured5 burnups of 61 - 98% of the initial 235U atoms. The results4»6-8 are excellent. Five miniplates wit UA^-AU hLE l fuel fabricate CNEe th A y witb d h U/cmg 1 3. hav- 0 e 3. also beeburnua o nt irradiatedpR estimatedOR e th n i * 3 to be as high as 90%1 . 5PIE are scheduled during 1984. ANL has fabricated two MEU UA12-A1 miniplates with about 3.0 g U/cm3 for irradiation testing in the ORR beginnin 1984n i g . Full-Sized Elements The following table lists the full-sized elements with MEU and LEU UAlx-Al fuel that have been irradiate beine ar gr o dIrradiate varioun i d s reactors. Also includeaverage th e ear d U burnup thabees ha tn achieved as of October 1983, th e235 goal average burnup, and the approximate dates for completion of the irradiations and the PIE. The data were obtained from the references thashowne ar t d thes,an e references shoul consultee b d r fo d further details.

39 UA1X-A1 Full-Sized Elements Approx. Approx. Ave. 235U Goal Irrad. PIE Fabri- Burnup, t Ave. 235U Cotnpl. Compl. cator Reacto_ r 10/83 Burnup, Z Date Date Ref«. MEU

CERCA SILOE l 2.2 50 50 1981 1983 9.10.11 ORR 3 1.7 56-76 50-75 1983 1984 9.10.12 JRR-2 2 1.6 0 40 1985 1985 13,14

7 1. NUKE 2 M ORR 59,73 50,75 1982 1984 6,12,15 4 1. 0 1 FRJ-1 50 45-50 1984 15,16,17 FRJ-2 10 1.1 5 45-55 1984 15,16,17 FRG-2 10 1.4 60 60 1984 15.16 4 1. 1 1 FRM 20 1985 15,16 6 1. 5 ASTRA 25 65 1986 15,18 6 1. SAPHI 5 R 15 65-70 1986 15,19,20 JMTR 2 1.6 0 60 1985 1985 13,21 LEU

CERCA HFR 2 2.1 48,73 50,75 1983 1984 9,22,23,24 ORR 2 2.1,2.3 0,27 50,75 1984 1985 9,12,22 2 2. * 1 JRR-4 0 40 1985 1985 13,14 1 2.2 0 1987 13,14

NUKEM IAE-R1 5 1.8 15 15,25

*T e irradiateb o JRR-2n di .

Whole-Core Demonstrations 3 A whole-core demonstration with LEU UA1X-A1 fuel and 1.7 - 1.8 g U/cm bega Decemben ni ForW M d2 re Nuclea 198th n 1i r Reactor Universite (FNRth t )a y of Michigan standar3 4 e Th .d fuel elements were manufacture CERCy b dNUKEd Aan M contro1 1 e anth dl fuel elements were manufacture NUKEMy b d . Since that time a substantial databas experimentaf o e l results 26s bee»2ha 7n accumulated an d several analyses resulte th 28 "f 3o 0s have been completed. Some experiments31 on LEU cores and mixed cores are still in progress.

Whole-core demonstrations with MEU UAlx-Al fuel and 1.6 g U/cm are planned SAPHIW M 0 1 R e ireactonth r (Switzerland JMTW M d Ran 0 5 e th n 1985n )i i d ,an 3 10 MW JRR-2 reactors (Japan) in 1986. A whole-core demonstration with LEU JRR-W M UA1 5 4U/cmg X -A3. reactoplannes 2 e 1i 2. th fue d n ri dlan (Japan) in 19873 .

3.2.3 U308-A1 Fuel

Small Test Plates PIE have been completed with excellent results'*"7 >mini6 323 ~n 3o -S plates fabricate ORNLy b d ,e CNENUKEMth Ad usinan , g UaOs-Al fue% 27 l - wit 0 2 h enrichment and 2.3 - 3.1 g U/cm3. All of the plates had been irradiated4 in the ORR to measured5 burnups of 80 - 98%.

40 Four LEU UßOg-Al mlnlplates fabricated36 by the CNEA with 3.1 - 3.6 g U/cm have been irradiated in the ORR to an average burnup estimated to be as high3 as 90%. PIE are scheduled during 1984. 5

Five miniplates fabricated by ORNL with MEU fuel and 2.5 - 3.1 g U/cm3 and one MEU miniplate fabricated by NUKEM with 21. .U/cmg 4 3 were also measuredo t irradiatedR OR e 5 th 90%burnup- n 11 *i 6 . if no s the ORR to measured5 burnups of 61 - 90%. Several of the ORNL plates with ~2.8 - 3.1 g U/cm3 and burnups greater than x 102 21 fissions/cm3 exhibited blisterin excessivr go e swelling .2 ThusU ,ME UßOg-Al fuel is not recommended for applications requiring fission densities approachin x 10 g2 2 1 fissions/cm^. It is worthwhile noting that the physical limit on fission density with U/cmg LE 1 s U10x i abou3. 3 fue 5 2t 1a 1. lt fissions/cmr fo 3 burnups approaching 100% of the 235U.

Juln I y 1983 smalo ,tw l test plates fabricate NUKEy b d M with UßOs-Al fuel, 40% enrichment, and 2.34 g U/cm3 began irradiation16 in a closed instrumented e FRJ-th loo n 2i p reacto t KFa r A Juellch, f SeptembeFRGo s A . r 1983, thed yha reache a dburnu f abouo p t 20%. Full-Sized Elements full-sizee Th d elementU LE d s an wit U hME fuel that have been irradiated or are being irradiated are shown below. referencee Th s shoule db consulte further fo d r details.

UjOg-Al Full-Sized Eleaenta Approx. Approx. 235, Av•»* e*TW W Goa«rWBlA IrradA»E » *»*PI •. * * IbM Fabri- No. of Burnup ,Z Ave . U Compl. Compl. caCor Reactor Els. 10/83 Burnup 235 .t Date Date Refs. MEU

TI ORR 1.7 55-72 50-75 1982 1984 12,35

UKAEA PLUTO 1 1.1 68 70 1982 1982 1,2.37 PLUTO 1 1.1 0 70 1984 1985 2.37

LEU

NUKEM DR-3 3 2.7 51-57 50 1983 1984 3,15 RFR 2 2.1 45,74 50,75 1983 1984 15.23,24 ORR 2 2.3 0,22 50,75 1984 1985 12,15 ASTRA 3 2.9 22 60 1986 15.18 FRG-2 7 3.1 8 65 1985 16,38 FRJ-2 6 2.9 0 50-60 1985 16,17 BER-2 2 2.0 0 1986 15,16

CERCA ORR 2 3.2 57,57 50,75 1984 1985 9,12,22 UKAEA PLUTO 1 2.9 0 70 1985 1985 2,37 FRJ-2 5 2.9 0 50-60 1985 2,16,17

41 3.2.4 U3SJ2-A1 Fuel

Small Test Plates

Four LEU U3S±2~A1 miniplates fabricated by ANL with about 3.75 g U/cm3 have been successfully irradiated in the ORR to a measured burnup of over initiae th U atomsf o l % . 90 Non-destructive examinations have been completed l fouonal r plate destructivd san e examinations have been f completeo o tw n o d the plates. The results S - * are excellent. The two plates that were not destructively examined were39 reinserte1 1 d into the ORR in March 1983 and irradiated tmeasureda o 5 burnu U3Si2~Af f abouo po t tse 95%lw miniplatene .A s witp hu to 5.6 g U/cm3 have been fabricated at ANL and began irradiation in the ORR in March 1984.

In July 1983, five small test plates fabricate NUKEy b d M wit U3Si2~AU hLE l fue uraniud lan m densitie 4.7f 5.0d so an 5 4 g/cm3 began irradiationn a n 1i 5 instrumented test loop in the FRJ-2 reactor. In September 1983 they had reache burnua d f abouo p t 20%. Full-Sized Plates Four full-sized plates being fabricated CERCy 2b 2 A wit U3Si2~AU hLE l fuel and 2.0, 3.5, 5.2, and 5.5 g U/cm3 are scheduled to begin irradiation11 SILOe ith n E reacto 1984n i r . Full-Sized Elements Six full-sized LEU U3Si2~Al elements with 19 plates, 0.51 mm meat, and 4.7 U/cmg 5 have been fabricate NUKEM,y b d . CERCA, Babcocd an & k Wilcox1*2 (B&W3 irradiatior )fo NUKEo ntw testinMORRe e elementTh th . n i g s bega6 15n 22 irradiatio CERCo tw y 1982 Ae Ma th elementn ;ni s began irradiatio Aprin ni l 1983; and the two B&W elements began irradiation in November 1983. NUKEe th Mf o element e On s reache averagn a d e burnun i abouf po % t41 January 1983. Results of the PIE to date are excellent. One of the CERCA elements reache averagn a d e burnu 50-55f o p Octoben %i expectee rar 1983E PI d. to begin in June 1984. The second NUKEM and the second CERCA element are expected to reach their goal average burnup of about 75% in August 1984, with E expecte PI e completeb W element o B& t de mid-1985th y b d sf o reache e On .n a d average burnu 50-55f po Aprin %i expectee lar 1984E begio PI t d. Novemben ni r 1984. The second B&W element is expected to reach an average burnup of about 75% around October 1984, with PIE expected to be completed by the end of 1985.

25 In February 1984, NUKEM fabricated four full-size U 3SidLE 2-Al elements with 23 plates, 0.51 mm meat, and ~3.7 g U/cm3 for irradiation test- FRG-e th in2n i greacto Geesthachtt a r , FRG. These element expectede sar 16 to reach an average burnup of about 65% by February 1986. Planning is in progress for irradiation testing in the R2 reactor (Sweden) of LEU U3Si2-Al elements fabricated by B&W, CERCA, and NUKEM. These elements would contain 18 plates, 0.76 mm meat, and uranium densities up to 4.8 g/cm3.

Whole-Core Demonstration Studies are in progress for a whole-core demonstration in the ORR with LEU U3Si2~Al fuel (~4.8 g U/cm3 and 340 g 235U per 19-plate element) beginning Sprine th f 1985n go i . CurrenU elementHE R tOR s wit plate9 h1 s contain about 28 2355g U.

42 The purposes of this demonstration are: (1) to gradually replace fresh HEU fuel with fres manneU fuea hLE n li whicn ri h most reactor conversions are expecte tako t d e comparo placet ) ,(2 e measurement transitiof so n core characteristics with calculations, and (3) to compare measurements of LEU equilibrium core characteristics with calculations. In addition, the experiments are expected to demonstrate that U3Si2~Al fuel elements fabricated under commercial conditions will perform reliabl normar fo y l reactor operation. About 100 elements are expected to reach the normal ORR burnup of 50% by the end of 1986. 3.2.5 U3SJ-A1 Fuel

Small Test Plates Fifteen LEU UßSi-Al miniplates fabricated35»39 by ANL with 4.8 and U/cg 7 m5. have beemeasureo t nR irradiateOR de th 96%- burnup n 4 i .3 d f o s PIE of the 4.8 g U/cm miniplates have been completed and the results12»1*0»1*1 U/cg 7 arm5. eplatee ver th schedulee yf ar s o good E PI 1984n .i d . Some possibilit improvemenr fo y t exist d furthean s r developmen tprogressn i wor s i k .

Three LEU U3Si-Al miniplates fabricated36 by the CNEA with 5.2 and 6.1 U/cg m hav eaveragn a bee o t n eR irradiate burnuOR e th p n i estimated e b o t d a sschedulee higar 90%s E a h PI . d during 1984.

UßSi-Af o t se l w miniplateAne U/cmg 2 7. s hav o witt ep hu bee n fabricated by ANL3 and the first plates began irradiation in the ORR in March 1984. NUKEM has fabricated six UßSi-Al miniplates with 6.9 g U/cm for irradiation testing in the ORR as part of this series. 3

In July 1983, three small test plates fabricated by NUKEM with LEU U3S1-A1 fue6.0d lan 5 U/cg m began irradiatio instrumenten a n i n d tese t th loo n i p FRJ-2 reactor Septemben I . r 1983 thereached ha y burnua d abouf po t 20%.

Full-Sized Plates Four full-sized plates wit U3Si-AU hLE l fuel fabricated CERCy 2b 2 A (two with 5.5 g U/cm3 and two with 6.0 g U/cm3) have been irradiated11 in the SILOE reacto averagn a o t r e 54%- maximue burnu2 Th .5 f po m- 70% 4 burnu6 . s pwa The thicknesses of these plates were measured periodically underwater in the pool between irradiation cycles using a special device to determine if swellings acceptabln wera n i e e range. Full-Sized Elements Based on these results, CERCA intends to fabricate o o a full-sized element with LEU UßSi-Al fuel and 6.0 g U/cm3 for irradiation testing11 in the SILOE reactor during 1984. The geometry of the plates and element will be identical with that of a standard SILOE element, and the total fissile mass will be about 510 g. Current SILOE HEU elements contain 330 g 235U.

3.2.6 USiAl-Al Fuel

Thirty-six LEU USiAl-Al miniplates fabricated35»39 by ANL with 4.5 - 7.0 g U/cra3 have been irradiated1* in the ORR to measured5 burnups of 34 - 96%. Since some of the plates with high burnup exhibited excessive swelling » * this fuel is no longer being pursued as a candidate for use in plate-typ40 e1 1fuel element geometries.

43 3.2.7 U6Fe-Al Fuel

ANs fabricateLha d fouUgFe-AU rLE l rainiplate sU/cmg wit r 8 fo h37- irradiatiobeginninR OR e th 1984 n i gn i n . However, o planthern t thie a se ar s time to pursue further the development of this fuel.

NUKEM plan fabricateo t s 16 >**x smalsi 3 l test plates usin U UgFe-AgLE l fuel with about 7.7 g U/cm3 for irradiation testing in an instrumented loop in the FRJ-2 reactor beginning in 1985. Five similar test plates with 5.5 g U/cm3 are planned to be irradiated in an instrumented loop in the FRG-2 reactor, also beginning in 1985.

3.2.8 U02 Caramel Fuel

The MTR OSIRIS reactor located at Saclay, France, has been operated successfully**'*"'** r ovefo >r four years wit whole-corha e loadin% 7 f go enriched UC>2 Caramel fuel developed by the French CEA. The reactor was loaded with Caramel fuel in October 1979 and the first power cycling began in February 1980. The first core loadin provisionaa d gha l maximum average burnu 20,00f o p 0 MWd/t. The permissible average burnup was increased to 30,000 MHd/t in February Septembef o d en e rth 1981 1983 t 2 elementA .23 , d beesha n irradiated,f o 1 *119 * which had reached an average burnup greater than 20,000 MWd/t. Out of these 191 elements, 140 had reached burnups between 28,000 and 30,000 MWd/t, which corresponds to maximum local burnups of between 40,000 and 43,000 MWd/t. Two elements had attained 40,000 MWd/t with local maxima of the order of 55,000 MWd/t. Improvements are possible1*1* in several aspects of Caramel fuel. These include reducin thicknese th gcurrene th f so t 1.4 meatm m 5 , increasine th g permissible average burnup possibl higs a 60,00s ho a t y 0 MWd/t, reducing costs by fabrication on an industrial scale and utilizing a dry process instead of the currently uset procesdwe conversior U02o d improvinsfo t an >g UF f no g procedure cuttinr fo s d chemicaan g l treatmen e fueth l f assemblieo t n si reprocessing whic technologyR bases PW hi n o d .

3 STATU3. PIN-TYPF SO E FUEL TECHNOLOGY

Substantial experienc s beeeha n obtaine AECy b d L (Canada fabricae th n )i - tion and irradiation behaviour of U-A1 alloy fuel at uranium densities up to 0.97 g/c . mSuc hdesigne fuelth f ,o s show bees Tabln ha ni n , esuccessfull3 y used in the NRU and NRX reactors over the past 20 years. More recently, fabrication, irradiatio d corrosionan n datbeine ar a g obtaine USiAl-Aln o d , .UgSil —A and U-A1 alloy fuel at higher uranium densities as part of a programme to evaluate the use of 20% EU in these reactors. In addition, investigations are in progres possible th f pin-typo £ fue n UU n o lighse li e U us teLE water SLOWPOKE-2 reactors.

developmene U-ZrHf o Th e us x pin-typd an t e fuel TRIGr fo s A reactors have been underway at GA Technologies since 1957. With only a few exceptions for high power or high usage facilities, TRIGA reactors have always used LEU. The original standard TRIGA-LEU fuel has a uranium density of 0.5 g U/cm (8.5 wt% U) and fuel with 0.7 g U/cm3 (12 wt% U) has been proved through successful reactor operation for over a decade. Previous work on U-ZrHx fuel during the SNAP reactor program had developed the technology up to 1.3 g U/cm found an indicatioo n d) U % wt f thi0 no (2 s bein limita g . earle Ith n y 1970s desire ,th r longe fo e r fuel lifetime increaseo t d le s s in uranium enrichment to 70% and 93%» In 1976, GA undertook the development

44 f o orden e i o allous o f) C r U e fuel Z wCh wt s U/cg 5 containin7 (4 m3. o C p gu LEreplaco C U highle Ch e y enriched fuels while maintaining long core life. Highly enriched versions of TRIGA fuel were discontinued by GA in January 1979

3.3.1 USiAl-A Ud 3Si-Alan l Fuels The USiAl-Al fuel material has been identified by AECL as the prime contender for both NRÜ and NRX such that both the current fuel- element designs and U densities can be retained. Investigation of the physical properties, fabricability, corrosio d irradiationan n characteristic underway.s i s 1* Mini-elements with a uranium density of 3.2 g/cm (Al-63 wtZ USiAl; 20Z EU) are being irradiate U burnu abouo t dpZ undet90 r conditions similao rt 3 those experience U fueNR ly b drods . Further irradiation progresn i e sar f o s mini-elements with a uranium density of 4.5 g/cm (Al-72 wt% USiAl; 202 EU). prime Ch Thi es i scandidat e replacement fueNRXr lfo . Alternative alloy compositions are also being investigated (U-3.2 wtZ Si, 3.0 wtZ Al as compared alsd effece an oth ) alternativf to Al Z wt 5 1. et ofabricatio, U-3.Si Z 5wt n methods. Since experience with plate-type fuels has shown Chat UßSi-Al has better high-burnup swelling resistance than USiAl-Al, the irradiaCion of rodded UßSi-Al fuel has been added to Che programme. 1C is now planned, commencing in 1984, to irradiate several complete rods of USiAl-Al and USi-Al fuel in NRU. 3.3.2 U-A1 Alloy Fuel

Similar irradiations to that for the USiAl-Al fuel have been corapleced. Mini-elements with a uranium density of 1.4 g/cm 3 (AI-37 wtZ U; 45Z EU) have bee n235Z burnua 1982n irradiate80 o Ui t f po U . NR n i d An NRX fuel rod has also been fabricated using this alloy. In this rod, only the central (0.95 m) section of each of the seven elements is fuelled. Thi ss currentl i fue d lro y being irradiate s schedulei d e removeb an d o t d d from the reacto 1984n i r . This fue back-ua ls a coulNRUr t pfo onlt ac d,s bu a y an interim back-up for NRX as higher uranium densities are required.

3.3.3 U02 Fuel

AEC currentls Li y investigatin possiblr lighn gfo i e t eus wate r SLOWPOKE-2 reactors pin-typU E Z ,20 2 fue eUÛ lwell-provee baseth n o d n CANDU power reacCor technology, but of substantially smaller (4.2 mm) fuel diameter.

3.3.4 U-ZrHx Fuel

A TRIGA-LEU fuel cluster has been undergoing irradiation51"53 in the ORR since December 1979 e standarTh . d geometry 16-pin cluste s containerha U LE d fuels with uranium loadings of 20, 30, and 45 wtZ U in several configurations. e activTh e fuediameten lpi Incoloe 12.9s th i r 0 , claddin5y80 mm g thickness si 0.41 mm, and the fuel length is 559 mm. In June 1982, pins with 20 wtZ U (1.3 g U/cm3) and 30 wtZ U (2.2 g U/cm3) were removed froclustee th m r after having reached their initial target average burnup values53 of 35Z and 40% of the contained 235U. The cluster was then reconfigured to contain nine pins with 45 wtZ U and seven dummy (unfueled) pins, and the irradiation was continued. One pin with 45 wtZ U was later removed for destructive burnup analysis and replaced with an unfueled pin. Subsequent preliminary analyses5 indicate that Che average burnup per pin of the four pins U range Z witwt s 0 h2 betwee n 47-59 d thae average Zan th th t f o e n burnupi r ppe four pins with 30 wtZ U ranges between 23-53Z (five are greater than 42Z and one was removed tese midwa alloo th t n i yw introductio w instrumentene a f no d pin). Peak radial averaged burnup highee factoa ar s y rb abouf ro t 1.25.

45 As of October 1983, the pins with 45 wt% U had been irradiated for up to 0 ful89 l powe d reacherha dayd dan s their initial target average 2 burnu50 f po of contained U with no indication of any problems. Preliminary estimates **

235 5 indicate that Uth averag235 e e burnup per pin of the seven full-test pins ranges from 55% to 65% of the contained 235U (two pins with lesser irradiation time containee havth f eo averagd% 23544 U)d ean . burnup % Pea33 f ko s radial averaged burnups are higher by a factor of about 1.25. In November 1983, the least sevee burneth nf o dfull-tes t pins failed under irradiatio mannea n ni r indicative of cladding failure followed by water logging of the fuel. The cause of this failure is still under investigation. During 1984, it is planned to reinsert the cluster back into the ORR for additional irradiation. As part of the development of the 45 wt% U fuel, GA has successfully completed a number of investigations that include out-of—pile measurements on fission product retention, quench testing d therma,an l cyclin pulsd an g e test- TRIGe th inAn i gDiegn Mar Sa kreactoF s o it laboratory t a r .

REFERENCES

In these references followino tw e ,th g proceeding f internationao s l meetings will be referred to by the shorter designations that are indicated:

ANL/RERTR/TM-4 (November 1982) Proceedings of the International Meeting on Research and Test Reactor Core Conversion FuelsU LE so t , froANL/RERTR/TM-4U mHE , CONF-821155, Argonne National Laboratory, Argonne, Illinois, November 8-10, 1982.

JAERI-M 84-073 (May 1984) Proceeding Internationae th f so l Meetin Reducen o g d Enrichmenr fo t Research and Test Reactors, 24-27 October 1983, Tokai, Japan, JAERI-M 84-073 (May 1984).

. SinclairD Char. P 1.d an e , "Developmen Enrichmenw Lo f to R FuetMT t la Dounreay," IAEA Semina Researcn ro h Reactor Operatio Used nan , IAEA-SR-77, Juelich, Federal Republic of Germany, 14-18 September 1981.

2. D. Sinclair, "Development of Low Enrichment MTR Fuel at Dounreay - Progress Report," JAERI-M 84-073 (May 1984).

Haack. K . ,3 TesU "IrradiatioLE td Fuean lU " ME Element 3, f no R D n i s JAERI-M 84-073 (May 1984). 4. J. L. Snelgrove, R. F. Doraagala, T. C. Weincek, and G. L. Copeland, "Fuel Development Activities of the U.S. RERTR Program, " JAERI-M 84-073 (May 1984). 5. J. L. Snelgrove (ANL), Personal Communication, March 1984. Burnup data are base masn o d s spectrographic measurement U isotopi f so c abundanc before and after irradiation. Detailed results are to be published. 6. M. F. Hrovat and H. W. Kassel, "Recent Status of Development and Irradiation Performance for Plate Type Fuel Elements With Reduced U Enrichment at NUKEM," JAERI-M 84-073 (May 1984). 7. E. Perez, C. Kohut, D. Giorsetti, G. Copeland, and J. Snelgrove, "Irradiation Performance of CNEA UA1X and U308 Miniplates," JAERI-M 84-073 (May 1984).

46 8. T. Shibata, et al., "Release of Fission Products From Irradiated Aluminide Fue Higt la h Temperature," ANL/RERTR/TM-4 (November 1982). Fanjas. Savorni. R B . Y . ,9 d nan "CERC A Contribution RERTe th Ro t sProgra m- Statu Developmentf so , November 1982," ANL/RERTR/TM-4 (November 1982).

10. F. Merchie, C. Baas, and M. Ploujoux, "Qualification in the Reactor SILOE Enrichew ofLo d Fuel Researcr fo s Tesd han t Reactors," ANL/RERTR/TM-4 (November 1982). Merchi. 11. BaasF C . d ,an e "Progress Repor Qualification o t Reactoe th n ni r Enrichew SILOLo f Eo d Fuel Researcr sfo Tesd han t Reactors JAERI-," M 84-073 (May 1984). 12. J. L. Snelgrove and G. L. Copeland, "Irradiation Testing of Full-Sized, RERTR Reduced Enrichment Fuel Elements," JAERI-M 84-073 ,(May 1984). 13. K. Sato, "Opening Statement to the International Meeting on Reduced Enrichment for Research and Test Reactors," JAERI-M 84-073 (May 1984). 14. M. Morozumi, H. Sakurai, J. Tsunoda, and Y. Suzuki, "Progress of Reduced Enrichment Progra JRR-n mi JRR-4,d 2an " JAERI-M 84-073 (May 1984). 15. M. F. Hrovat and H. W. Hassel, "Recent Status and Future Aspect of Plate Type Fuel Element Technology With High uranium Densit NUKEM,t a y " ANL/RERTR/TM-4 (November 1982). Thamm. G . ,16 "The GermaStatue th f nso AF-Program," JAERI-M 84-073 (May 1984). 17. G. Thamm (KFA Juelich), Personal Communication, April 1984.

18. J. Casta (ÖFZS), Personal Communication, March 1984. 19. H. Winkler and J. Zeiss, "Comparison of Calculations and Measurements of ME SAPHIUe Fueth n Rli Reactor," ANL/RERTR/TM-4 (November 1982). 20. H. Winkler (EIR), Personal Communication, April 1984. Nakayama21. F . . Itabash!,Y Kanekawa. ,H Oyamada. ,R Saito. ,M d ,an T. Nagamatsuya, "Progress of Reduced Enrichment Program in JMTR," JAERI-M 84-073 (May 1984).

22. Y. Fanjas, Ph. Dewez, and B. Savornin, "CERCA Contribution to the RERTR Progra mStatu- Developmenf o s Septembet- r 1983," JAERI-M 84-073 (May 1984). 23. H. Pruimboom and R. J. Swanenburg de Veye, "Status Report on the Irradia- TesU tioLE tf no Element Pettee th n nsi High Flux Reactor," ANL/RERTR/TM-4 (November 1982). Pruimboom. H . 24 Lijbrink. ,E Otterdijk n Swanenbur. va J . . ,R K d Veye,an e gd , "Status Report on the Irradiation Testing and Post-Irradiation Examination of Low-Enriched u^Og-A UA1d lan X-A1 Fuel Element Netherlande th y b s s Energy Research Foundation (ECN)," JAERI-M 84-073 (May 1984). Hassel. W Muelle. H . H ,. d r25 an "NUKE M MTR-Newslette Decembe" 8, . rNo r 1983. 26. D. K. Wehe and J. S. King, "FNR Demonstration Experiments - Part I: Beam Port Leakage Current Spectra,d an s " ANL/RERTR/TM-4 (November 1982).

King. S . , J Weh. "FNd K ean . D R Demonstratio . 27 n Experiment Pars- : II t Subcadmium Neutron Flux Measurements," ANL/RERTR/TM-4 (November 1982).

47 28. J. A. Rathkopf, C. R. Drumra, W. R. Martin, and J. C. Lee, "Analysis of the Ford Nuclear Reacto Core,U rLE " ANL/RERTR/TM-4 (November 1982). 29. M. M. Bretscher and J. L. Snelgrove, "Comparison of Calculated Quantities with Measured Quantities for the LEU-Fueled Ford Nuclear Reactor," ANL/RERTR/TM-4 (November 1982).

Arigan. K Tsuchihashi. K d . an e30 , "Analysi Criticaf so l ExperimentR FN f o s LEU Core," JAERI-M 84-073 (May 1984). Lee . DruramWeheC . . KingMartin. . K R S ,R . 31 . J . ,D . . C ,.d ,J W ,an "Operating Experience, Measurement, and Analysis of the LEU Demonstration FNR,e Coreth "t sa JAERI- M 84-073 (May 1984). 32. G. L. Copeland and J. L. Snelgrove, "Examination of Irradiated High-U- Loaded UßOg-Al Fuel Plates," ANL/RERTR/TM-4 (November 1982). Copeland. L . G . ,33 "The Aluminum-UsOs Exothermic Reaction," ANL/RERTR/TM-4 (November 1982) Posey. C . J , "Releas. 34 Fissiof eo n Products from Miniature Fuel Platet a s Elevated Temperature," ANL/RERTR/TM-4 (November 1982). 35. J. L. Snelgrove, "RERTR Program Progress in Qualifying Reduced-Enrich- ment Fuels," ANL/RERTR/TM-4 (November 1982). GiorsettPerez. . R E . . E D ,d "Argentin. ian 36 e Activities Relatee th o t d Developmen Enrichew Lo f o t d Fuel Elements," ANL/RERTR/TM-4 (November 1982).

37. R. Panter (Harwell), Personal Communication, April 1984. 38. W. Krull (GKSS), Personal Communication, March 1984. Domagala. F Thresh. . Weincek. R R C . . . H , T ,39 d ,an "U-S U-Si-Ad ian l Disper- sion Fuel Alloy Developmen Researcr fo t Tesd han t Reactors," ANL/RERTR/TM-4 (November 1982). 40. G. L. Hofraan, L. A. Neimark, and R. F. Mattas, "Irradiation Behavior of Experimental Miniature Uranium Suicide Fuel Plates," ANL/RERTR/TM-4 (November 1982). Neiraark. A . L d ,an "Irradiation ma f Ho . L n. G Behavio . 41 Uranium-Suicidf ro e Dispersion Fuels," JAERI-M 84-073 (May 1984). 42. J. B. Freira, "Overview of Babcock & Wilcox Involvement in the RERTR Program," JAERI-M 84-073 (May 1984). 43. W. Krull, S. Nazaré, and E. Wehner, "Necessity for Additional Irradiation U MiniplateTestLE f so Irradiation i s n Devices," German AF-Program Internal Report, January 1984. 44. M. Barnier and J. P. Beylot, "OSIRIS, a MTR Adapted and Well Fitted to LEU Utilization: Qualification and Development," JAERI-M 84-073 (May 1984).

Cerles. M . J Trotabas. , M . 45 DeContenson. ,G DeLafosse. J d ,an , "FrencU hLE Fuel for Research Reactors With Emphasis on the OSIRIS Experience of Core Conversio Reactod nan r Operatio Fuel,w n Ne wit e "th h IAEA Seminan ro Research Reactor Operation and Use, IAEA-SR-77, Juelich, Federal Republic of Germany, 14-18 September 1981.

48 46. F. Cherruau, "The Caramel Fuel in OSIRIS: The Complete Conversion of a High Flux Research Enrichew ReactoLo a o rt d Fuel," Proceedinge th f so International Meetin Developmentn go , Fabricatio Applicatiod nan f no Reduced Enrichment Fuel Researcr fo s Tesd han t Reactors, ANL/RERTR/TM-3, CONF-801144, Argonne National Laboratory, Argonne, Illinois, November 12-14, 1980. Feraday. A . Foo. DavidsonWinegar. . M T D E . ,. ,. . M R J 47 d ,, an "The Thermal Stability of AL-USiAl Dispersion Fuels and Al-U Alloys," Nuclear Technology _5£3 (Augus23 , t 1982).

48. J. C. Wood, M. T. Foo, and L. C. Berthiaurae, "The Development and Testing of Reduced Enrichment Fuel Canadiar fo s n Research Reactors," ANL/RERTR/TM-4 (November 1982).

WoodFooBerthiaurae. . . T C C . . . ,, M J L Schaefer. Herbert. D N . . . J 49 ,L d ,an , "Advance Manufacturine th n i s Irradiatiod gan Reducef no d Enrichment Fuels for Canadian Research Reactors," JAERI-M 84-073 (May 1984). Burbidge. A Hilbor . . W G . d J ,nan . "SLOWPOKE50 Firse Th :t Decad Beyond,d ean " AECL-8252 (October 1983).

Gietzen. J . A . , 51 "Improved Capabilit U-Zrf o y H Fuels," Proceedinge th f o s International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuel Researcr sfo Tesd han t Reactors, ANL/RERTR/TM-3, CONF-801144, Argonne National Laboratory, Argonne, Illinois, November 12-14, 1980. . CheswortWest. H B . . R ,G d "Statu. han 52 Developmenf so Anticipated an t d Performanc Uranium-Zirconiuf o e m Hydrid Fuel,U eLE " IAEA Semina Researcn ro h Reactor Operation and Use, IAEA-SR-77, Juelich, Federal Republic of Germany, 14-18 September 1981. 53. G. B. West, "Current Status of U-ZrH LEU Fuel Experiments in the ORR," ANL/RERTR/TM-4 (November 1982). 54. G. B. West (GA) and J. L. Snelgrove (ANL), Personal Communication, March 1984.

49 A-1

ADDENDUM APPENDIXA

Enrichment Reduction Calculations for Heavy Water Type Reactors

Matos . FreeseE . Komoriy E . . H J . ,d K ,an a

Reduced Enrichment Research and Test Reactor (RERTR) Program

Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439 U.S.A.

ABSTRACT Neutronics calculations are presented on the feasibility for conversion, from HEU fuel to MEU and LEU fuels, of three research and test reactors (DIDO, DR-3, and JRR-2) using heavy wate moderatos a r d coolantran . Detailed results include uranium densities, burnup performance d neutro,an n flux performance for each reactor.

51 A-2

ADDENDUM APPENDIXA TABLE OF CONTENTS Page A.1 Introduction ...... A-4

A.2 Methods and Procedures ...... A-4 A. 2.1 Methods ...... A-4

A.2.2 Procedures ...... À-4

A.3 The DIDO Reactor...... À-5

À.3.1 Design Specifications ...... 5 .À-

A. 3.2 Calculational Models ...... A-5

A.3.2.1 Cross Sections ...... 5 .A- A.3.2.2 Cold Clean Core Model ...... 5 .A- A.3.2.3 Axial Extrapolation Lengths...... 5 .A- A.3.2.4 Fuel Management Strategy ...... A-10 A.3.2.5 Cole Burnuth d f Cleapo n Core ...... A-10 A.3.2. Loadg Operationad Ri 6san l Poisons ...... A-10

A.3.3 Performance Results for HEU(75%) Reference Core ...... A-12

A.3.4 Performance Results with MEU(45%) Fuel ...... A-12

A.3.5 Performance Results with LEU(20%) Fuel ...... A-17 A.3.6 Reactivity Worth Fissiof so n Products ...... A-20 A.3.7 Reactivity Worth Naturaf so l Boron Poisons ...... A-20

A.3.8 Conclusions ...... A-23 A.4 The DR-3 Reactor ...... A-24

Â.4.1 Design Specifications ...... A-24

A.4.2 Calculational Models ...... Â-24

A.4.2.1 Cross Sections ...... Â-24 A.4.2.2 Cold Clean Model ...... A-24 A.4.2. Loadg Operationad Ri 3san l Poisons ...... A-24 A.4.2.4 Axial Extrapolation Lengths ...... A-29 A.4.2.5 XY Model and Fuel Management Strategy ...... A-29

A.4.3 Performance Results for HEU(93%) Reference Cores ...... A-29

A.4.4 Performance Results with MEU(45%) Fuel ...... A-36

52 A-3 Page A.4.5 Performance Results with LEU(20%) Fuel ...... A-36

A.4.6 Flux Ratios Inside Central Thimble of Fuel Element ...... A-40

A.4.7 Reactivity Worths of Fission Products ...... A-40 A.4.8 Reactivity Worths of Natural Boron Poisons ...... A-40

A.4.9 Conclusions ...... Â-44 JRR-e A.Th 25 Reactor ...... A-45

Â.5.1 Design Specifications ...... A-45

Â.5.2 Calculational Models ...... A-45

A.5.2.1 Cross Sections ...... A-45 A.5.2.2 Axial Extrapolation Lengths ...... A-45 A.5.2.3 Model for Burnup Studies ...... A-51

A.5.3 Performance Result HEU(93%r fo s ) Reference Core ...... A-51

A.5.4 Performance Results with MEU(45%) and LEU(20%) Fuels .... A-51 A.5.4.1 Results witFueU hME l ...... A-51 A.5.4.2 Results wit FueU hLE l ...... A-56

A.5.5 Conclusions ...... A-56

53 A-4

A.I Introduction

Heavy-water-moderated research and test reactors using HEU fuel currently utilize a number of fuel element designs. These designs can be organized into three general categories: (1) concentric tube designs, as characterized by the DIDO class reactors, (2) MTR plate-type designs as characterized by the JRR-e th 2y b reacto d HFBan ) othe Japann A (3 i rR US d r reacto e an ,designs th n i r , F reacto RH X reactor e NR th rd s characterizey Canadn an a b i s U d NR aan e th y b d in France.

This appendix summarizes the methods and procedures used at ANL for computing the feasibility for conversion of these types of reactors, and presents the results of detailed calculations for the DIDO reactor in the UK, the DR-3 reacto Denmarkn JRR-e i r th 2d reacto,an Japann i r . Detailed results include uranium densities with MEU and LEU fuels, burnup performance, and neutron flux performanc eacr fo eh reactor.

2 MethodA. Procedured an s s

Method1 A.2. s

The methods and codes used at ANL for analysis of heavy-water-moderated research and test reactors are similiar to those described for light-water- moderated reactor IAEAppendie n si th A f Guidebooo xA Researcn ko h Reactor Core Conversion U FuelLE so st fro(IAEA-TECDOC-233U mHE , August 1980).

2 AProcedure.2. s The calculational procedure used for analysis of each reactor can be summarize followss a d :

Generat• e cross section functioa corr s fo sea r burnuf fo no d pan reflector.

Hex-r o Z Z XY mode p u colf lo t dSe clea• n core.

• Compute axial extrapolation lengths by least-squares-fit of fluxes.

modeY X burnur p fo lu t p Se studie • choosd san fueea l management strategy based on information provided by reactor operator.

• Incorporate natural boron to approximate reactivity losses due to rig loads and operational poisons.

Comput• e end-of-cycle (EOC) excess reactivity, 235jj burnup distributio d fluxe an nequilibriur fo s U referencmHE e core with cycle length specifie y reactodb r operator.

• Compute uranium densities needed to match cycle length and EOC excess reactivity of HEU reference core using MEU and LEU fuels. • Compare burnup and flux performance of MEU and LEU cases with HEU reference case.

54 A-5

A. 3 The DIDO Reactor A.3.1 Design Specifications The design specifications used in the ANL calculations for the DIDO reactor are shown in Table Al. These specifications were obtained through correspondenc personad ean l communication with Harwell personnel.

A.3.2 Calculational Models A.3.2.1 Cross Sections Five-group microscopic cross sectioncore th e r werfo s e generatea s a d function of burnup using the EPRI-CELL code. The DIDO fuel element was modeled as shown in Fig. Al with all four fuel tubes and the outer aluminum wrapper tube explicitly defined. An actual DIDO fuel tube has axial tapers extending about 31 mm from each end, and different clad thicknesses on the inside and outside of the fuel meat. For calculational purposes, Harwell and ANL personnel agree modeo t dreference lth U fueeHE l element using fuel meat wit uraniua h m density of 0.65 g/cm^, a total of 205 g 235^ and a uniform tube thickness of entire th r e fo 609. 1.6m m m axia86m l height. Wit coolana h t channel thickness of 3.09 mm, these assumptions lead to a fuel tube model with an average meat thickness of 0.72 mm, and an average clad thickness of 0.48 mm. All calculations were performed without boro gadoliniur no m burnable oute poisoth -n no side of the aluminum wrapper tube. The thickness of the ring of heavy water outside the wrapper tube was chosen to preserve the volume of the heavy water associated with each fuel element heavl Al .y wate s assumerwa havo t d ea light water impurity of 0.25%. Separate cross sections were also preparereflectore th r fo d s usin homoa g - genized core source and appropriate thicknesses of heavy water and graphite. A.3.2.2 Cold Clean Core Model A horizontal cross section of the DIDO reactor is shown in Fig. A2. The core consist 5 fue2 lf s o element 4-6-5-6-a n i s 4 configuration e cors Th .ewa modele geometrZ XY n i d y (Fig wit ) frese . A3 hth U fue hHE l element heavd san y water moderator represented separately. The temperature of thé fuel plates coolane s 20° th d tha wa moderatod Can f o tan t s 25°rwa C since heavy water cross section 20°t a s C wer availablet eno . Control rods g loadsri ,d an , operational poisons were not included at this stage. The aluminum and heavy elemene wateth d boxen i ren t s were homogenized with 0.25 volume fraction aluminu 0.7d an m5 volume fraction heavy water excese Th . s reactivite th f o y cold clean cors compute ewa 23.4%e b o t d. A.3.2.3 Axial Extrapolation Lengths Axial extrapolation lengths were calculated for the cold clean HEU core usin least-squares-fia g firse e fluxemesw th th fe tr h o fo st interval s Z modelabov XY midplane e th e. th f Sinco e e microscopic cross sections with burnup corresponding to the middle of the equilibrium cycle (MOC) are to be use subsequenn i d t burnup calculations mino,a r adjustmene th s mad twa o t e fresh core extrapolation length yielo t se sam th de excess reactivity using cross section r botfo sh zer oburnue burnuth d p an pcorrespondin C MO o t g conditions. The adjusted extrapolation lengths used in all subsequent calculation geometrY X n i s y wit U fueshowe hHE lar Tabl n n i . AxiaeA2 l extrapolation lengths calculated in a similar manner for each region of the MEU and LEU cores are also shown in Table A2.

55 A-6

Table Al. DIDO Reactor - Description of Design Parameters CalculationL UseAN n i d s Reactor Design Description

Reactor Type Tank, DIDO Type Steady-State Power LevelW ,M 25.5 Number of Standard Fuel Elements 25 Numbe Controf ro l Fuel Elements None Irradiation Channels Cente Eacf ro h Fuel Element Core Geometry 4-6-5-6-4 Configuration Lattice Pitch, mm? 2 15 x 2 15 Active Core Volume, A 354 Core Average Volumetric Power Density, kW/A, 72 Average Linear Power Density, W/cm 3937

Moderator, Coolant D20 Reflector Û20, Graphite Burnup Status of Core Equilibrium Core

Fuel Element Design Description

Fuel Type Four Concentric Fuel Tubes plus Unfuelled Aluminum Wrapper Tube uranium Enrichment,% 75 Outer Tube O.D.m ,m 103.0 Plate Thicknessm ,m 1.68 Coolant Channel Thickness, mm 3.02.2d an 99 (outermost channel) Fuel Meat Material U-A1 Alloy Average Fuel Meat Thicknessm ,m 0.721

Fuel Meat Dimensionsm ,m Tube 1: R0=31.2905, Ri=30.5695, L=609.6

Tube 2: R0=36.0605, Ri=-35.3395, L=609.6

Tube 3: R0=«40.8305, Ri=40.1095, L=609.6

Tub : R4 e o=45.6005, Rj.~44.8795, L-609.6 Clad Material Aluminum Average Clad Thickness, mm 0.4795 Uranium Density in Fuel Meat, g/crn^ 0.65 235u/Standard Fuel Element,g 205 Average 235u Discharge Burnup,% 40-50

56 A-7 Figur . DesigAl e DIDf no O Fuel Element.

\ \ \

Geometry Data Component Thickness, mm .Cell Volume Fractions Fuel Tubes*(4) 1.68 Fuel Meat 0.0299 Outer Al Tube 3.13 D20 0.8879 Coolant Channels 3.09 Al 0.0822 Fuel Meat 0.721 Clad 0.4795 U CorHE e *Fueled Lengt h- 609. m 6m 235U/Element 205 g U Density 0.65 g/cm3 U Enrichment 75%

57 A-8

Figure A2. Horizontal Cross Section of the DIDO Reactor.

GRAPHITE l CONTROHH D MO L

CONCRETE

ION CHAMBER TUBE

Sin.nOHTUART HOLE(I)

MOLE1 in.0 )S(1

STEEL TANK Jin-MORTUARY HOLE{l) SHIM-SHUT Off SHEETS (DOWN) AlUHINIUM TANKS FUEL ELEMENT HORIZONTAL SECTION DIDO REACTOR

58 A-9

Fig3 DIDA . O Reacto rHorizonta- l Cross Sectio Cort na e Midplane Z ModeXY . l UseCalculatior fo d Colf no d Clean CorExtrapolatiod ean n Lengths ModeY X . l User fo d Burnup Studies.

Graphite Reflector

D20 Reflector Tank

D-O Moderator

-Fuel Elemen snmirn . 52mm

Tabl . ExtrapolatioeA2 e th r fo n Length) mm n s(i HEU, MEU, and LEU Cores

0.51 mm 0.7 Meam 2m t Meat HEU MEU LEU LEU Graphite Reflector 546.6 507.3 508.9 506.4 Aluminum Tank 467.2 428.5 430.2 427.5 D2Û Reflector 297.6 272.8 274.7 271.5 Û2Û Moderator 227.7 219.0 221.3 216.9 Fuel Elemen1 t 215.6 215.8 217.5 221.0 2 222.9 222.0 223.7 227.6 3 203.3 205.5 207.1 209.9 4 209.6 210.8 212.4 215.6 5 224.8 223.6 225.3 229.4 6 199.2 202.1 203.6 206.3 7 201.9 204.4 205.9 208.7 8 211.8 212.7 214.3 217.6

59 A-10

A. 3.2.4 Fuel Management Strategy Operational data on typical fuel element changes for 13 consecutive new element loading patterns between June 1979 and May 1980 for the 26 element PLUTO reactor were provided to ANL by Harwell. About eight elements are typicall ydowy replaceda n 4 , up d y wit da eacf o h4 h2 d fres en e h th fue t la operation cycle. Analysis of this fuel replacement pattern indicated that, in general dividee b core n th ,ca e d into three concentric ring calcular fo s - tional purposes. Elements near the center of the core are typically replaced r threo ever o etw ycycles . secone Thosth n dei rinreplacee ar g d about every three cycles, and those near the edge are replaced every three or four cycles.

This informatio 5 elemen2 s adaptee nwa th to t DIDd O reacto develoo t r p a fuel management strategy based on an average three reactor cycle life for each fuel element n thiI .s model e fueth ,l elements were divided inte th o three group selement8 showd an Fign ni , swit, 8 . beinA4 , h9 g replaced after consecutive operational cycles. The average 235U discharge burnup of DIDO fuel 50%o elementt .0 4 s i s

A.3.2.5. ColBurnue th d f Cleapo n Core

e purposTh thif eo s calculatio determino t s nwa 235e th eu burnup distribution of the EOC equilibrium core so that natural boron poisons could then be incorporated intmodee th o o represenlt t e reactivitth o t e ydu losseC EO t sa rig loads, the residual burnable poison, and the cold-to-hot reactivity swing. As mentioned in Section A.3.2.2, the excess reactivity of the cold clean core was computed to be 23.4%. Using the REBUS-3 fuel cycle analysis code in XY geometry, fuel element loadings for the EOC equilibrium core were established by repetitive burnup calculations witfuee th hl replacement pattern described above. Equilibrium operatio s attainenwa d after seve excesC n EO cycles d s an reactiG , BO wite -th h vities repeating in groups of three in subsequent cycles. For cycle 10 of the depleted core, the EOC excess reactivity was computed to be 10.4%. The reactivity loss due to fuel depletion was 6.3% and that due to all of the fission products was 6.7%.

A.3.2.6 Rig Loads and Operational Poisons An EOC reactivity balance table provided by Harwell was then used as a basi modelinr fo s reactivite th g y in-coro losset e d reflectosdu ean g rri loads, residual burnable poison, and the cold-to-hot reactivity swing. To model these EOC reactivity effects, natural boron was added to the experiment centee spacth f eact o ra e h fuel elemendepletee th n i t d core o (cyclt ) e10 reduc excese th e s reactivit 5.5y yb represeno %t n in-cora t loag 4.5f eri o d % an residuaa d l burnable poison . wortNatura1% f ho l boro s alsnwa o uniformly homogenized in the heavy water reflector to reduce the excess reactivity by 2% to represent the reflector rig load. The excess reactivity was further reduced by 1.4% by adding natural boron to the coolant channels of each elemen represeno t cold-to-hoe th t t reactivity loss. Thus, having incorporated a total poison wort 8.9%f ho excese ,th s equilibriuC reactivitEO e th f o ym core after 10 cycles was 1.5%.

These natural boron concentrations were then fixed for all subsequent calculations with HEU, U fuelsMEULE d ,.an

60 A-11

Fig. A4. DIDO Reactor - XY Model, and Three Cycle Life Fuel Management Schem Burnur fo e p Studies

293.77 .

236.93 - 235.80 -

184.89 - 2 1 1 2 169.69 - 1 3 3 3 3 1 154.49 - 2 2 1 2 2

1 3 3 3 3 1 124.09 - 2 1 1 2 108.89 - V 57.97 - Tank 56.84~ Graphite

0.0 1 J I I I I I I I I I I I 1 I J 1 N O N O N O N O N O N O N O o» N O O NN O S O O o co ON C MS CO«TNO\OCC c o O v O J - O C M CO ON • • * o r«. •-•CO in vO O o COON m m ON es CN es

61 A-12

A.3.3 Performance Results for HEU(75%) Reference Core

e resultburnue Th th f pso calculatioU coreHE e , th repeate r fo n d witl al h natural boron poisons present e showar ,Tabln n i . ThieA3 s table includes: ke efth f ) value (1 functioa s seven-cycle sa th f timr no fo e e approaco t h the equilibriu r thremfo cored cycleean s after equilibrium operatio s beenha n reached repea0 1 e kd n thiei ff'tTh an . cycler s , s9 fo grou , s8 f thre po e s in subsequent cycles eacf ke o e ff'Th hd . en fou shuty e listeda rth - t a d down include the reactivity increases due to both the Xe decay and the fresh fuel replacement; (2) the number of fuel elements replaced after each four day shutdown; (3) the fresh element 23^u loading and the uranium density in the kC eff'HO d san averageG BO e dfueth ove) threle (4 th meatr d e an ;consecutiv e cycles afte e equilibriuth r m cor s beeeha n achieved shoult I . notee b d d that the EOC excess reactivity varies from cycle to cycle in the equilibrium core (cycles differenceo 8-10t e )numbee du th d location an rsi elemente th f no s thareplacede ar t .

The 235U burnu pafteC distribution EO cycle0 1 rd an sG (27BO t 6a s days)

are shown in Fig. A5. The nine elements to be discharged and replaced with fres nexe hth t fue r cycl lfo markede ar ee averag Th .U core burnu235 e th e n i p s computes 32.1%wa e wa e 17.0 n FigG b th C I a .BO to , EO d tha t d. % A6 an t a t average 23^u burnup of the fuel elements discharged from each core position after three consecutive cycles of equilibrium operation are presented. The average discharge burnups averaged over all elements in the core is 45.4%.

Fast, epithermal, and thermal fluxes at EOC for a midplane traverse along the x-axis through the center of the core are shown in Fig. A7. The peak and the average thermal fluxes in the core were computed to be 4.46 x 3.6d an 110*x ^ 1*10 n/cm2/s, respectively heave th yn I wate. r reflectore th , peak thermal flu s 4.1xwa 0x 10* 1* n/cm2/s.

A.3.4 Performance Results with MEU(45%) Fuel

Wit U fuelhME , burnup studies were performed usin same th ge methods, fuel element geometry, and natural boron poison concentrations as for the HEU core. Wit fixeha d cycle lengtdays4 2 f h,o several calculations (includina g four day shutdown) were first done with different 235U loadings. The 235U loading that approximately matched the average EOC excess reactivity of the HEU equilibrium core was found by interpolation. Another calculation was then obtaio t n nru fluxe burnud san p distributions. The reactivity history, element loading, uranium density, and average BOG and EOC keff's for the equilibrium core are shown in Table A3. The required uranium density was computed to be 1.13 g/cm3, which corresponds to a 235 235C C UUEO EO loadine d contenth an 214.G d f elementr o gBO an tpe 4e g Th . Pu conten eacn i t h fuel elemen showe ar tFign ni 9 afte.A cycle0 1 r s (276 days). The average 235U burnup in the core at BOG was computed to be 16.0% s 30.2%anwa e averagdC . Th EO tha t a te 235U fuee burnuth l f elementpo s discharged from each core position are shown in Fig. A10. The average discharge burnups averaged ovel element core al r 43.0%s th i e n i s .

62 A-13

Tabl . DIDeA3 O Reacto Functiokr- a e fs fa Timf HEUr no fo e ,U LE MEU d an , Cores with Three Cycle Lifetime Management Scheme and AveragU 235 e Loadings, Uranium Densities C kEquilibriud eff'EO ,an d san G BO m Number of Meam m 0.7t2 Elements Replaced keff (75) keff (45) keff (20) Approac Equilibriuo t h m Core 1 0 25* 1.2059 1.1903 1.1696 1.1770 24 1.0930 1.0833 1.0707 1.0750 2 28 9 1.1582 1.1463 1.1309 1.1362 52 1.0449 1.0400 1.0339 1.0358 3 56 8 1.1300 1.1208 1.1089 1.1128 80 1.0126 1.0115 1.0104 1.0104 4 84 8 1.1358 1.1266 1.1139 1.1185 108 1.0173 1.0163 1.0145 1.0152 5 112 9 1.1294 1.1209 1.1097 1.1135 136 1.0114 1.0112 1.0108 1.0107 6 140 8 1.1231 1.1152 1.1050 1.1083 164 1.0046 1.0053 1.0062 1.0056 7 168 8 1.1339 1.1251 1.1130 1.1174 192 1.0151 1.0146 1.0134 1.0139

Equilibrium Core 196 9 1.1288 1.1206 1.1095 1.1132 220 1.0108 1.0108 1.0106 1.0105 9 224 8 1.1229 1.1151 1.1049 1.1083 248 1.0044 1.0052 1.0061 1.0055 10 252 8 1.1338 1.1251 1.1130 1.1174 276C 1.0150 1.0145 1.0134 1.0139

235U/Element, g 205.0 214.4 229.3 219.6 p£5» S/cm3 0.4875 0.5099 0.5453 0.7379 3 , n g/cmP 0.650 1.1330 2.7264 3.6897

Averag kG efeBO f 1.1285 1.1203 1.1091 1.1130 for Cycles 8, 9, and 10

Averag kC efeEO f 1.0101 1.0102 1.0100 1.0100 for Cycles 8, 9, and 10

aFresh core witloadg operationad ri h an s l poisons. 0 repea1 bkgff' d thin i tan cycler s, 9 fo sgrou , s8 f thre po subsequenn ei t cycles. used for flux performance comparison and U loading distributions. 235

63 A-14

Fig. DIDA5 . O Reacto U (75%HE r- ) Fuel, 0.7 2Meam m t Thickness

BOC and EOC Distribution of 235U and EOC Distribution of Pu

a U Enrichment: 75% BOC k ff: 1.1338 U Density: 0.65 g/cm3 EOC ke": 1.0150b Fresh Fuel Loading: 205 g 235U Cyclee£ength: 24.0 Days r 235U (BOC) 171.6 139.1 139 .1 171.6 235U (EOC) 143.1 113.4 113.4 143.1 Pu (EOC) 0.81 1.24 1,24 0 81 r r 235U Contents 144.0 205.0 205.0 205.0 205.0 144.0 119.9 169.3 165.6 165.6 169.3 119.9 Fresh: 5125.0g 0.99 0.64 0.72 0.72 0.64 0.99 BOC: 4253.4g EOC: 3482.3g 168.8 164.7 r!27.6 164.7 168.8 138.2 130.9 98.9 130.9 138.2 1.02 1.24 1.63 1.24 1.02 235 144.0 205.0 205.0 205.0 205.0 144.0 Average U Burnup 119.9 169.3 165.6 165.6 169.3 119.9 BOC: 17.0% 0.99 0.64 0.72 0.72 0.64 0.99 EOC: 32.1% 171.6 139.1 139.1 171.6 143.1 113.4 113.4 143.1 0.81 1.24 1.24 0 81

fres8 d han elementse X BO Co n cors .eha

EOC corequilibrius eha m fission product concentrations.

Fig. A6. DIDO Reactor - HEU (75%) Fuel 235U Discharge Burnup CO.7 Meam 2m t Thickness) Numbe Cyclef ro s Burned 3 3 3 3 235U Discharge Burnup, % 42.4 44.7 44.7 42.4

3 3 3 3 3 3 41.5 45.4 49 0 49.0 45.4 41.5

3 3 3 3 3 45.4 49.8 51.8 49.8 45.4

3 3 3 3 3 3 41.5 45.4 49 .0 49 0 45.4 41.5

3 3 3 3 42.4 44.7 44. 7 42.4

64 A-15

Fig. A7. DIDO Reactor FluxeC - EO t Corsa e Midplane HEU Fue Meam m l t2 Wit0.7 h

Q. o O o 0 i f \ / O q _ fast *" n ... epithermal ._ thermal

o d 0.0 12.5 25.0 37.5 50.0 62.5 75.0 87.5 100.0 112.5 125.0 137.5 150.0 X - AXIS (cm)

Fig. A8. DIDO Reactor FluC - EO x Ratio Cort sa e Midplane MEU Fuel with 0.72 mm Meat

*45'*75

o 8- o CM i_ O O o m C. x

»O o -«• •^^ X«

_ fas_ t ._ epithermai .__ thermal

^ 25.0tZ 0 0. 37.5 5O. ^ O75.62 0 8 7J KKXO 112^ 125.0 137J 1SaO - XAXI S (cm)

65 A-16

. FigDIDA9 . O Reacto U (45%ME r- ) Fuel, 0.7 Meam 2m t Thickness

DistributioC EO d BO an GDistributioC EO 235f d no Uan u P f no

U Enrichment% 45 : BOG k a eff 1.1251 U Density: 1.13 g/cm3 EOC k b eff 1.0145 Fresh Fuel Loading: 214.4 g 235U Cycle Length: 24.0 Days

235U (BOG) 181.7 149.5 149.5 181.7 235U (EOC) 153.6 123.8 123.8 153.6 Pu (EOC) 1.87 2.88 2.88 1.87 r r 235U Contents 154.4 214.4 214.4 214.4 214.4 154.4 130.3 179.3 175.7 175.7 179.3 130.3 Fresh: 5360.0 g 2.32 1.43 1.63 1.63 1.43 2.32 BOG: 4502.9 g EOC: 3739.1g 178.9 174.8 137.9 174.8 178.9 148.5 141.2 108.9 141.2 148.5 2.34 2.84 3.81 2.84 2.34 Average 235U Burnup 154.4 214.4 214.4 214.4 214.4 154.4 130.3 179.3 175.7 175.7 179.3 130.3 BOC: 16.0% 2.32 1.43 1.63 1. 63 1.43 2.32 EOC: 30.2%

181.7 149 .5 149.5 181.7 153.6 123 .8 123.8 153.6 1.87 2.88 2. 88 1.87

BOC core has no Xe and 8 fresh elements. EOC core has equilibrium fission product concentrations.

Fig. A10. DIDO Reactor - MEU (45%) Fuel 235U Discharge Burnup (0.7 Meam 2m t Thickness) Numbe Cyclef ro s Burned 3 3 3 3 235U Discharge Burnup,% 40.0 42.3 42.3 40.0

3 3 3 3 3 3 39 3 43.0 46.5 46.5 43.0 39.3

3 3 3 3 3 42.9 47.2 49.3 47.2 42.9 3 3 3 3 3 3 39 .3 43.0 46.5 46.5 43.0 39.3

3 3 3 3 40 0 42.3 42.3 40.0

66 A-17

Fast, epithermal, and thermal flux ratios at EOC between the MEU and HEU cases are shown in Fig. A8 for a midplane traverse along the x-axis through the epithermae th coree cente e fasd th Th .an tf ro l fluxes wit U fue e nearlhME lar y unchanged from those of the HEU reference. In the core, the peak and the average thermal flux ratios are 0.91 and 0.92, respectively. At the peak in the heavy water reflector thermae th , l flux rati 0.95s i o shoult .I notee b d d from Figs9 .A and A5 that the ratio of the core 235u contents at EOC between the MEU and HEU case 0.93s si . This vervalun i ys ei goo d agreement witcomputee th h d core- average thermal flux rati 0.92f o .

A.3.5 Performance Results with LEU(20%) Fuel

Using the same fuel element geometry and natural boron poison concentra- U core HE calculatione e th , th n tioni s a ss describe previoue th n i d s section U fueonME l were repeated usin fuelU LE g .

The reactivity history, element loading, uranium density, and average BOC equilibriue kC aneth ff'EO d r sfo m core (cycles 8-10 showe )ar Tabln ni . eA3 e requireTh d uranium densit s compute ywa 2.7e b o 3t d g/cm3 , which corresponds t235a o u loadin elementr 229.f go pe 3g . 235C EO u contenP u d C conten an eacn EO i t C d hBO an t fue e Th l element are shown in Fig. All after 10 cycles (276 days). The average 235u burnup in s 27.7%s computewa thwa e e 14.6C eb C Th . EO d BO o thacor t d%t an t a ta e average 235u burnup of the fuel elements discharged from each core position is shown in Fig. A12. The average discharge burnups averaged over all elements in the core is 39.4%.

Fast, epithermal, and thermal flux ratios at EOC between the LEU and HEU casemidplana showe r sar fo Fign ni 3 e A1 .travers e alon x-axie th g s through the coree centefase e ratioth Th th t. f f o rso (>5.5 3epithermae th keV d )an l fluxes averaged over the core are 1.01 and 0.97, respectively. In the core, thaverage eth pead ean k thermal «0.62 flu) 5eV x ratio 0.8e 0.83d ar s0 an , respectively heave th peae yn th i k wate t A . r reflector e thermath , l flux ratio is 0.90. From Figs. All and A5, the ratio of the LEU and HEU EOC core 235u contents is 0.84. Again, this is in very good agreement with the computed core-average thermal flux ratio of 0.83*

Influenc Thickef eo r Fuel Meat

If the clad thickness can be reduced from 0.48 mm to 0.38 mm as in many other light water and heavy water reactors, the fuel meat thickness can be increased from 0.72 mm to 0.92 mm without altering the fuel tube thickness and without significantly alterin steady-statee th g , thermal-hydraulic safety margins. Since the fuel meat volume would be increased by a factor of 1.28, the uranium density correspondin 235g elemenr loadina 9 pe uo 22 t g f to g woule b d about 2.1 g/cm3. The uranium densitites for clad thicknesses between 0.38 mm and 0.48 mm can be estimated by interpolation. However, the co-extrusion process and the Harwell preference for a tapered core end would have to be re-examined to see whether all the theoretical density reduction can be achieved. Flux and burnup performance are expected to be very similar to the values quoted for the above LEU fuel case since performance is determined primarily 235e bth yu loadin r elemene platpe gth f ei t thicknes e sameth .s i s

67 A-18

Fig. All. DIDO Reacto U (20%LE r- ) Fuel, 0.7 Meam 2m t Thickness

BOG and EOC Distribution of 235U and HOC Distribution of Pu

EnrichmentU % 20 : 3 BO* G cc k ' 1.1130 efi f Ü Density: 2.73 g/cm3 b EOC ks.ef _,.f ; 1.0134 235 Fresh Fuel Loading: 229.3g U Cycle Length: 24.0 Days 235U (BOG) 197.5 r!66.0 r!66.0 197.5 235U (EOC) 169.6 140.3 140.3 169.6 Pu (EOC) 3.95 6.13 6.13 3.95 r r —— 235U Contents 170.5 229.3 229.3 229.3 229 .3 170.5 Fresh: 5732.5 g 146.3 195.3 191.8 191.8 195.3 146.3 BOG: 4898.2g 4.99 2.93 3.32 3.32 2.93 4.99 EOC: 4145.3 g 194.9 191.1 155.0 191.1 194.9 165.0 158.1 125.9 158.1 165.0 4.87 5.90 8.07 5.90 4.87 235 r 3 •* , Average U Burnup 170- s 229 £.^22-• *.Q7 229 • J 229 * «J 170.5 146.3 195.3 191.8 191.8 195.3 146 .3 BOG: 14.6% 4.99 2.93 3.32 3.32 2.93 4.99 EOC: 27.7% 197.5 166.0 166.0 197.5 169 .6 140.3 140.3 169 .6 3.95 6.13 6.13 3.95

BOG core has no Xe and 8 fresh elements. EOC core has equilibrium fission product concentrations.

Fig. A12. DIDO Reactor - LEU (20%) Fuel 235U Discharge Burnup (0.72 mm Meat Thickness) Number of Cycles Burned 3 3 3 3' 235U Discharge Burnup, % 36.9 38.8 38.8 36.9

3 3 3 3 3 3 36.2 39.3 42.5 42.5 39.3 36.2 3 3 3 3 3 39.4 43.2 45.1 43.2 39.4 3 • 3 3 3 3 3

36.2 39.3 42.5 42.5 39 .3 36.2 3 3 3 3 36.9 38.8 38.8 36.9

68 A-19

. DIDAl3 O . ReactoFig r- EOC Flux Ratios at Core Midplane FueU Meam LE lm tWit2 h0.7

g. V 1 L. o k. oO o o /—N"u?«1 x^_ -* §• ' .••••-....•• _ . IL. .••""'-. •x^ *

O g *•••" -_ — _ • * »• ^ ^ ^ s >T s Z3 "> / _«j . . / _ fas_ t o •'\.r*~'' epithermal 03 f .__ thermal Ô~

r»

0 12.0. 5 25.0 37.0 5150. 50.5 0137. 62. 0 5125. 75.5 0112. 87. 0 5100. X - AXIS (cm)

. DIDA14 O . ReactoFig r- EOC Flux Ratio Cort sa e Midplane FueU Meam LE lm tWit1 h0.5

*20/*75

- 9 o r"_ « a. »— L. o o O CM o O o

O. x___ rJ

m m o

m *•""" ~ ~~ ' ^ ™" "* ™ ^ ••

« r^''^-' __ fast O ..... epithermal CD' .__ thermal 0~ in

0.0 12.5 25.0 37.5 50.0 62.5 75.0 87.5 100.0 112.5 125.0 137.5 150.0 X - AXIS (cm)

69 A-20

Influenc Thinnef eo r Fuel Tubes

An additional calculation with LEU suicide fuel was performed in order to determin e increas th e therma th n i e l flux that coul obtainee b de th f i d fuel tubes coul made b d e thinner e thickeTh . r coolant channels would provide more heavy wate neutror fo r n thermalization e casTh .e studied here used fuel meat wit n averaga h e thicknes d claddin 0.5f an so m 1m g wit averagn a h e thicknes 0.3f o s, thu 8mm s givin tuba g e thicknes f 1.2so 7m insteae m th f o d curren e thicknest Th coolan1.6e th . 8mm f so t channel s increaseswa d from 3.09 mm to 3.50 mm and the diameter of the experiment space at the center of each fuel elemen s increasetwa 0.4y db . 1mm

Burnup studies were performed usin same th ge methods, fuel management strategy, and natural boron concentrations as those for the HEU reference design. The reactivity history is shown in Table A3. The 235U loading that approximately matche average th d excesC eEO s U casreactivits HE ewa e th f o y compute 219.e b elementr o pe t d 6 g , which correspond uraniua o t s m densitf o y 3.69 g/cra3. The BOG and EOC 235U content and the EOC Pu content in each fuel elemen showe ar tFign ni afte 5 . A1 cycle0 1 r s (276 days) average Th . e U s s compute29.0% wa wa e 15.3 b C C .o BO corEO e burnud that d %t th t an ea a t n i p 235 e averagTh e 23^U fuee burnuth l f elementpo s discharged from each core position are shown in Fig. A16. The average discharge burnup averaged over all elements core i41.2%s th nei .

Fast, epithermal and thermal flux ratios at EOC between the LEU case with 0.5meam m 1 treferencU thicknesHE e th d e a s an cas show e r ear fo Fign ni 4 .A1 midplane traverse along the x-axis through the center of the core. The ratios of the fast (>5.53 keV) and the epithermal fluxes averaged over the core are 1.00.97d an 0 , respectively average coree th peae th d th ,n ekan I .therma l «0.625 eV) flux ratios are 0.85 and 0.87, respectively. At the peak in the heavy water reflector thermae th , l flux rati 0.91s i o n summaryI . , thinner fuel tube d thickean s r coolant channels would increas average eth e thermal fluU corLE xe e ratifroth n mi o 0.8 30.87o t . A.3.6 Reactivity Worths of Fission Products Individual reactivity worths of Xe, Sm, and the remaining fission products were calculated at EOC for th 135e HEU ,lif9 MEU, and LEU cases at 276 days. e resultTh s liste Tabln i d 4 shoeA w only small difference worthn si s amone th g four cores. A.3.7 Reactivity Worths of Natural Boron Poisons As mentioned above, the natural boron concentrations used to represent the reactivity losses due to rig loads, residual burnable poison, and the cold-to-hot reactivity swing corC werEO e e U onlcalculated HE an ye th r fo d were then fixed for subsequent calculations with MEU and LEU fuels. Table A5 shows that the reactivity worths of these poisons at EOC are smaller in the U coreU coreLE HE d MEse . an Uth tha n i n

70 A-21

Fig.AlS. DIDO Reacto U (20%LE r- ) Fuel, 0.5 Meam m 1 t Thickness

BOG and EOC Distribution of 235U and EOC Distribution of Pu EnrichmentU % 20 : BOG k 1.1174' eff U Density: 3.69 g/cm3 EOC k 1 eff 1.0139 Fresh Fuel Loading: 219. 235g 6 U Cycle Length: 24.0 Days

235U (BOG) 187.6 156.0 156.0 187.6 235U (EOC) 159.9 130.6 130.6 159.9 Pu (EOC) 3.76 5.80 5.80 3.76 235U Contents 160.8 219.6 219.6 219.6 219.6 160.8 137.0 185.1 181.5 181.5 185.1 137.0 Fresh: 5490.0 g 4.74 2.79 3.17 3.17 2.79 4.74 BOG: 4650.1 g EOC: 3897.3 g 184.8 180.8 144.5 180.8 184.8 154.9 147.7 115.7 147.7 154.9 4.63 5.60 7.60 5.60 4.63 r r^ Average 235U Burnup .8 160 219 .6 219.6 219 .6 219.6 160.8 BOG: 15.3% 137.0 185.1 181.5 181.5 185. 1 137.0 4.74 2.79 3.17 3.17 2.79 4.74 EOC: 29.0% f f 187.6 156.0 156.0 187. 6 159 .9 130.6 130.6 159.9 3.76 5.80 5.80 3.76

BOG core has no Xe and 8 fresh elements. 5EOC cor equilibrius eha m fission product concentrations.

DID6 FigA1 .O Reacto (20%U LE r- ) Fuel 235U Discharge Burnup (0.5 Meam m 1 t Thickness) Number of Cycles Burned 3 3 3 3 235U Discharge Burnup, % 38.4 40.5 40.5 38.4

3 3 3 3 3 3

37.6 41.2 44.6 44.6 41.2 37.6 3 3 3 3 3 41.2 45.4 47.3 45.4 41.2

3 3 3 3 3 3 37.6 41.2 44.6 44.6 41.2 37.6 3 3 3 3 38.4 40.5 40.5 38.4

71 A-22

Tabl . ReactiviteA4 y Worths (ok/kFissiof o ) ,% n Products in EOC cores with HEU, MEU, and LEU Fuels

0.72 mm Meat 0.51 mm Meat Fission Product Removed HEU MEU LEU LEU

Xe 3.47 3.51 3.54 3.53

S m 0.62 0.63 0.63 0.63

Lumped F. P. 2.61 2.47 2.28 2.38

Total 6.70 6.61 6.45 6.54

Table A5. Reactivity Worths (<5k/k, %) of Natural Boron Poisons in the EOC Cores with HEU, MEU and LEU Fuels

0.7 Meat______m m 2 _ 0.5m 1m Meat

Poison EOC EOC EOC EOC Removed HEU MEU LEU LEU In-Core Rig Loads 5.55 5.19 4.68 4.98 and Burnable Poison Reflector Rig Loads 2.03 1.92 1.89 1.87 r Natfo B . 1.40 1.31 1.18 1.26 Temp. Increase

Total 8.98 8.42 7.75 8.11

72 A-23

A.3.8 Conclusions The results of this study indicate that It Is technically feasible to convert the DIDO reactor to both MEU and LEU fuels without changing the core lifetime, the present rig loads, or the excess reactivity at end of cycle if suitable fuels are demonstrated, available, and licensable. No changes are required in the thickness of the fuel tubes and little or no changes are expecte e steady-stateth n i d , thermal-hydraulic safety margins. e followinTh g table summarize y resultke e th ss U witLE hd HEUan U ,ME fuels if no changes are made in the current fuel element geometry. All cases have a cycle length of 24 days, the same natural boron concentrations repre- g sentinload ri operationad se an th g same th U corele HE d poisone ,an th f so average end-of-cycle excess reactivity as the HEU core.

HEU(75%) MEU(45%) LEU(20%)

Uranium Density, g/cm3 0.65 1.13 2.73 235U Loading per Fresh Element, g 205.0 214.4 229.3 Average 235jj Discharge Burnup, % 45.4 43.0 39.4 Average Flux Ratios in Core Fast O5.53 keV) 1.01 1.01 Epithermal 0.99 0.97 Thermal «0.625 eV) 0.92 0.83 Thermal Flux Ratio at Peak in Heavy 0.95 0.91 Water Reflector

average Ith f e thicknes clae th d f scoulo reducee b d d froe currenth m t average th 0.3o t , 0.4 8em mm m 8thicknes fuee th l f measo t coul increasee b d d from 0.9o 0.7t withoum 2m 2m m t changin thicknese th gcurrene th f so t fuel tubes. Thus fuee th ,l meat volume coul increasee b d factoa y b df 1.28 ro . For the same ^U loadings per element as the MEU and LEU cases shown above,

the uranium densitie23 U fuel LE estimatee d ar ss abou e an b wit 9 U o 0. t hME d

3 1 g/cm2. ,g/cmd respectivelyan 3 . Uranium densitie intermediatr fo s e fuel mea d claan td thicknesse estimatee b n sca interpolationy b d . Flud burnuxan p performanc l case al expectee sar r aboue b fo esame o s thosth t da e e shown ni the table above. If the thickness of the fuel tubes could be decreased from 1.68 mm to 1.27 mm by utilizing fuel meat with an average thickness of 0.51 mm and cladding wit n averaga h e thicknes e thicknes th 0.3f coolane so , th 8 mm f so t channels coul increasee b d d from provido 3.5o 3.0t t m 0m 9m m e more heavy wate neur fo r - tron thermalization. The 235U loading required to match the EOC excess reac- presene tivitth U f coro ytHE e wit currens hit loadg tri s compute swa e b o t d 219.6 g per element, which corresponds to a uranium density of 3.69 g/cm3. The fast and epithermal fluxes ratios in the core and the thermal flux ratio heave th peae ayn th t i k wate r reflector woul virtualle b d y identical with those for the LEU case in the above table, but the average thermal flux ratio in the core would be increased from 0.83 to 0.87. With LEU fuel, a reduction of the average thermal flux in the core to 0.8currene 0.8- th 3 f 7o t value could require reoptimizatio certaif o n n irradiation rigs and the manner in which the reactor is operated.

73 A-24

DR-e Th 3A .Reacto4 r A.4.1 Design Specifications The design specifications used in the ANL calculations for the DR-3 reacto e showar rTabln ni . TheseA6 e specifications were obtained through correspondence and personal communication with RISO personnel.

A. 4.2 Calculât!onal Models A.4.2.1 Cross Sections Five-group microscopic cross sectioncore th e r werfo s e generatea s a d functio burnuf no p usin EPRI-CELe th g L codeDR-e Th 3. fuel elemen modeles twa d as shown in Fig. A17 with all four fuel tubes and the outer aluminum tube explicitly defined e thicknesTh .rine heavf th go f so y water outsid fuee th el element was chosen to preserve the volume of the heavy water associated with each fuel element heave Th .y wate s assumerwa havo t d ligha e t water impurity of 0.25%. Separate cross sections were also preparereflectore th r fo d s usina g homogenized core source and appropriate thicknesses of heavy water and graphite. A.4.2.2 Cold Clean Model A horizontal cross section of the DR-3 reactor is shown in Fig. A18. The core contains 26 fuel elements arranged in a 4-6-6-6-4 configuraton. The core was modelegeometrZ XY n i d y (Fig. A19) witfrese th U hfue hHE l elementd an s heavy water represented separately e temperaturTh . fuee th l f plateeo s 20°swa C an coolande th tha moderatod f to tan 25°s rwa C since heavy water cross sections at 20°C were not available. Control rods, rig loads, and operational poisons were not included at this stage. The aluminum and heavy water in the element end boxes were homogenized with 0.25 volume fraction aluminum and 0.75 volume fraction heavy water. The excess reactivity of the cold clean core was computed to be 22.2%. A.4.2. Loadg Operationad Ri 3san l Poisons Usin fuee th gl replacement strategy described below, equilibrium cycle fuel element loadings corresponding to a cycle length of 23.5 days were computed starting from the cold clean core in order to obtain a burnup distribution representative of EOC conditions with no rig loads. The EOC excess reactivit thif o y s depleted cor 9.54%s ewa . Natural boron was then added to the experiment space at the center of each fuel elemen represeno t C exces reduco % EO t 9 e y sb th e t reactivitd an % 4 y b y the in-cor g loaderi s suggeste RISOy e actuab d Th . l boron poison worths com- thesn i e e us putecalculation r fo d s were 4.15 9.23%d an % , respectively. The EOC excess reactivity for each rig load case was further reduced by about 1.4% to represent the cold-to-hot reactivity loss. This 1.4% reduction was represented usin componentso tw g reductioa ) (1 :abouf no t 0.7s wa % incorporated by adding natural boron to the coolant channels of each element to represent the estimated reactivity component due to spectral hardening, and (2) a reduction of about 0.7% was made by adjusting the axial extrapolation lengths (see Section A.4.2.4) to represent the estimated reactivity increaseo t los e sdu d leakage.

e naturaTh l boron concentrations computed above were thel n al fixe r fo d subsequent calculations with HEU U fuels,LE MEUd .an ,

74 A-25

Table A6. DR-3 Reactor - Description of Design Parameters CalculationL UseAN n i d s

Reactor Design Description

Reactor Type Tank, DIDO Type Steady-State Power LevelW ,M 10 Number of Standard Fuel Elements 26 Number of Control Fuel Elements None Irradiation Channels Cente Eacf o r h Fuel Element Core Geometry 4-6-6-6-4 Configuration Lattice Pitch2 ,mm 152. 152.x 4 4 Active Core Volume,i 360.1 Core Average Volumetric Power Density, kW/2. 28 Average Linear Power Density, W/cm 790 Moderator, Coolant D20

Reflector Graphit0 2 D e Burnup Statu Corf o s e Equilibrium Core

Fuel Element Design Description

Fuel Type Four Concentric Fuel Tubes plus Unfuelled Aluminum Outer Tube Uranium Enrichment, % 93 Outer Tube O.D., mm 103.12 Plate Thickness, mm 1.52 Coolant Channel Thickness, mm 3.38 Fuel Meat Material UA14-A1 Fuel Meat Thickkness, mm 0.50

Fuel Meat Dimensionsm ,m Tub R: O1 e=31.4 , Ri-30.9, L=6lO Tube 2: RO=O6.3, Rj-35.8, L-610 Tub R: 0e3 =41.2 , R^O.?, L=610 Tube 4: R0=46.1, R.^45.6, L-610 Clad Material Aluminum Clad Thickness, mm 0.51 235u Densit Fuen i y l Meat, 0.5083 235U/Standard Fuel Element,g 150 Average 2^^U Discharge Burnup, % 40-50

75 A-26

Figure Al7. Desig DR-f no 3 Fuel Element.

Geometry Data Cell Volume Fractions Component Thickness, mm Fuel Meat 0.0208 Fuel Tubes*(4) 1.52 0.9151 Oute Tubl rA e 1.57 0.0641 Coolant Channels 3.38 Fuel Meat 0.50 Clad 0.51 HEU Core 235U/Element 150 g 235U Density in Meat 0.508 g/cm3 *Fueled Length - 610 mm

76 A-27

Figure A18. Horizontal Cross Sectio DR-f no 3 Reactor

1in. STEEL TANK LEAD O.SIn.Al ION CHAMBER CRAPHITE IKSIOE 0.25in.BORAI BONDING TANK ^in.LEAD PLUS ASSEMBLY STORAGE HOLES REFLECTOR

ION CHAMBER FINE CONTROL SHIM SAFETY FUEL CONCRETE PLUS ASSEMBLY ROD PLATES RODS ASSEMBLY SHIELD (7) (2) (26)

77 A-28

Fig. A19. DR-3 Reactor - Horizontal Cross Section at Core Midplane of XYZ Model Used for Calculation of Extrapolation LengthModeY X d lan s Use r fo d Burnup Studies

Y A

Graphite Reflector

•Tank Reflector

"Fuel Element

'DO Moderator

X m m 2 15

78 A-29

A.4.2.4 Axial Extrapolation Lengths Using the results of the boron-poisoned calculations for the representa- tivcoreC eEO , axial extrapolation lengths were computed fro least-squaresa m - e firs e fluxeth mesw th fe t r fi ho fo t s tinterval s abov core th ee midplane. As mentione e previouth n i d s section extrapolatioe th , n e lengthth r fo s fuel elements were then adjusted to reduce the excess reactivity by about 0.7% to represen e estimateth t d componen cold-to-hoe th f to t reactivityo t lose sdu leakage. The resulting extrapolation lengths for each region of the HEU core are show Tabln i n. eA7 As also noted in the previous section, the same boron poison concentrations calculated for the HEU core were also used in the calculations with MEU and LEU fuels. However, the extrapolation lengths for the MEU and LEU cases (also shown in Table A7) were separately adjusted to yield a reactivity reduction of approximately 0.7%. A.4.2.5 XY Model and Fuel Management Strategy Operational dat fuen ao l element consecutivn changete r fo s e cycles (240- DR-e 249th 3f )reactoo r between October 197 Juld 9an y 1980 were provideo t d ANRISOy Lb . Typically, about four elements were replaced with fresh fuel after reaching their goal burnup of approximately 50%. The average cycle length over the ten runs was 23.5 days. Analysi fuee th l f sreplacemeno t patterns indicated thacore n th tca e be divided into three concentric rings (see Fig. A20 calculationar )fo l purposes. Generally, fuel elements near the center of the core are replaced about every five cycles. Those in the second ring are replaced every six cycles, and those in the outer ring are replaced about every seven cycles. The REBUS-2 fuel cycle analysi burnue s th cod uses r pewa fo d calculation s geometriY nX y (Fig. A19). Equilibrium cycle loadings were compute replaciny b d g appropriate th e fue th n li f 1/5o f eaco 7 ,eh d 1/ 1/6elementen d e an ,th t sa 23.cycley da 5 . Since there werelemento tw e s wit fiva h e cycle lifetimen te , elements with a six cycle lifetime, and fourteen elements with a seven cycle lifetime, approximately four full elements were replaced with fresh fuel during each loading change. A.4.3 Performance Results with HEU(93%) Reference Cores The results of burnup calculations for the HEU core with rig loads of 4.2% and 9.2% are summarized in Tables A8 and A9, respectively. For an average cycle lengt 23.f ho 5excesC daysEO e sth , reactivities were computee b o t d 4.1% wit4.2e th h loag -0.9 %d ri an d % wit9.2e loadg th h %ri . These results indicate that normal operating procedure modifiee b y nee so ma t d operato t d e contenu P 235C C UEO EO t d e contenwitan th G d 9.2a h BO an t loadg %e ri Th . in each fuel element are shown in Fig. A21 for the case with a 4.2% rig load. The average 235y burnup of the fuel elements discharged from each core positio shows i n Fign ni . A22 e averag.Th e discharge burnup averaged over all element core 48.9%s th ewa n i s. Fast, epithermal, and thermal fluxes for a midplane traverse along the x-axis throug core centee th th he f wito r 4.2a h loag showe %ri ar d n Fig ni . A23. average th e pead Th ean k thermal fluxe core s th e«0.62 n weri ) e5 eV computeo t d be 1.8 510*x ^ 1.5d n/cm^/s410 an 4x , respectively heave th yn I wate. r reflector peae th ,k thermal flu^ 1.5s n/cm^/s10 xwa 9x .

79 A-30

Table A7. Axial Extrapolation Lengths (in mm) for DR-3 Reactor with HEU FuelU , MEULE sd ,an

4.2% ôk/ Loag kRi d

Region HEUU LE MEU LEU (.76 mm meat)

Graphite Reflector 478.6 479.7 481.3 481.2 Aluminum Tank 448.6 449.8 451.5 451.4 D2Û Reflector 300.0 301.6 303.7 303.6 Û2Û Moderator 209.9 212.4 215.4 215.2 Fuel Element 1 187.1 188.6 189.7 189.6 2 194.2 195.5 196.5 196.4 3 174.6 176.1 177.3 177.2 4 180.1 181.6 182.7 182.6 5 194.6 195.9 196.9 196.8 6 171.3 172.8 174.0 173.9 7 176.1 177.6 178.7 178.6 8 189.4 190.9 191.9 191.8

9.2% ôk/k Rig Load

Region HEU MEU LEU

Graphite Reflector 479.5 480.4 481.4 Aluminum Tank 449.6 450.5 451.6 D2Û Reflector 301.0 302.4 303.5 Ü2U Moderator 211.0 213.3 215.0 Fuel Elemen1 t 188.9 190.5 191.0 2 196.0 197.5 197.9 3 176.4 178.0 178.5 4 181.9 183.4 183.9 5 196.4 197.9 198.3 6 173.0 174.6 175.1 7 177.9 179.4 179.9 8 191.3 192.8 193.3

80 A-31

Fig. A20. DR-3 Reactor - XY Model and Fuel Management Scheme for Burnup Studies

209.38- 208.25-

157.42-

7 7 7 7 142.18-

7 6 6 6 6 7 126.94- 7 6 5 5 6 7 111.70. 7 6 6 6 6 7 96.46- 7 7 7 7 81.22-

D20 30.39- 29.26_ Tank

Graphite

0.0-n n _L -,=i~*ro^^« J"

^^ • • « • •••• «•• « eoOi—icn^s-voo > c* o FH iH i-l iH i-t CM CM CM

81 A-32

Tabl . DR-eA8 3 Reacto 235- r U Loadings with Uranium Enrichmentd an % 45 f so Matco t Fuee % th hl 20 Cycle Exces C LengtEO d sh an Reactivitf o y Enriche% the93 d Reference Core Loag witabouf Ri o d h a t 4.2% ok/k Specifie RISOy b d a.

Enrichment, % 93 45 20 20

Fuel Type l -A 1 UA UA1 -Al UA1 -Al UA1 -Al X X X X Cycle Length, days 23.5 23.5 23.5 23.5

Meat Thicknessm ,m 0.50 0.50 0.50 0.76

Clad Thickness, mm 0.51 0.51 0.51 0.38

235U/Element, gC 150.0 156.0 165.6 164.3

P25, g/cm3 0.508 0.529 0.561 0.366

PU. S/cm3 0.547 1.175 2.806 1.831 Wt.% U(7 v/o void) 18.4 33.7 57.9 45.4

2H/235Ud 2236 2150 2025 2048

6 BOC keff 1.0929 1.0877 1.0816 1.0823

EOC keff 1.0431 1.0433 1.0430 1.0432 23 5U Burned, gf 73.3 72.4 71.2 71.2

DischargU 5 3 2 e Burnup,% 48.9 46.4 43.0 43.3

aThe HEU U case, havLE l MEUsam e sal d th e ean , natural boron concentrations representing the rig loads. Based on a power level of 10 MW. frese th C235hr feefo U wt.,d e d an ar P25 fue %, U lP , element. 2H/235U in fresh fuel element, including 34.42 mm thick moderation channel surrounding each element. SThe BOC and EOC cores contain equilibrium concentrations of Xe, Sm, and lumped fission products, and the BOC-EOC reactivity difference is burnuo t thae pdu t only. Averag dischargel al f eo d fuel elements.

82 A-33

Table A9. DR-3 Reactor - 235U Loadings with Uranium Enrichments of Matco t Fuee % hth l 20 Cycld an e Exces % C Lengt45 EO d shan Reactivity of the 93% Enriched Reference Core with an Alternativ Loag abouf eRi o d t 9.2% 6k/k Specifie RISOy b d a.

Enrichment, % 93 45 20

Fuel Type UA1 -Al UA1 -Al UA1 -Al X X X

Cycle Length, days 23.5 23.5 23.5

Meat Thicknessm ,m 0.50 0.50 0.50

Clad Thicknessm ,m 0.51 0.51 0.51

235U/Element, gc 150.0 155.7 163.5

p , g/cm3 0.508 0.528 0.554 25

3 pu, g/cm 0.547 1.172 2.770 voido v/ Wt.)7 U( % 18.4 33.6 57.5

2H/235Ud 2236 2154 2051

e BOC keff 1.0419 1.0364 1.0307

EOC keff 0.9914 0.9915 0.9916 235U Burned, gf 73.3 72.3 71.1

235U Dis charge Burnup, % 48.9 46.4 43.5

HEU, MEU, and LEU cases all have the same natural boron concentrations representing the rig loads. Basepowea . n o drMW leve0 1 f lo

frese th hr 235 feefo U ,e wt.d d ar an P25 fue %> U lP , element.

235 UH/ id2 n fresh fuel element, including 34.42 mm thick moderation channel surrounding each element. The BOC and EOC cores contain equilibrium concentrations of Xe, Sm, and lumped fission products BOC-EOe th d C,an reactivity differencs i e that due to burnup only. Averag dischargel al f eo d fuel elements.

83 A-34

Fig. A21. DR-3 Reactor - HEU (93%) Fuel, 4.2% Rig Load

BOG and EOC Distribtuion of 235U and EOC Distribution of Pu Base n Fueo d l cycle Length Matching Criterion (0.5 Meam 0m t Thickness)

U Enrichment: 93Z BOC k 1.0929 U Density: 0.55 g/cm3 . EOCk' 1.0A31 Fresh Fuel Loading: 150 g 23SU Cyclee£ength: 23.5 Days

(BOC) 114.1 112.2 112.3 114.1 (EOC) 103.5 101.2 101.2 103.5 Pij(EOC) 0.12 0.16 0.16 0.12 235U Content 115.2 117.3 114.6 114.6 117.3 115.2 Fresh: 3900.0 g 104.9 105.6 102.1 102.1 105.6 104.9 BOC: 2993.3 g 0.12 0.15 0.12 0.16 0.16 0.15 EOC: 2696.0 g 114.6 115.8 119.4 119.4 115.8 114.6 104.1 103.6 105.7 105.7 103.6 104.1 0.13 0.16 0.16 0.16 0.16 0.13 Average 235U Burnup 115.2 117.3 114.6 114.6 117.3 115.2 104.9 105.6 102.1 102.1 105.6 104.9 BOC: 23.2% 0.12 0.15 0.16 0.16 0.15 0.12 EOC: 30.9%

114.1 112.2 112.2 114.1 103.5 101.2 101.2 103.5 0.12 0.16 0.16 0.12

Fig. A22. DR-3 Reactor - HEU (93%) Fuel, 235U Discharge Elemenf o d BurnuEn t t Lifetimea p s

Burned 7 7 7 7 235U Discharge Burnup% nup, , 49.5 51.7 51.7 49.5

7 6 6 6 6 7 48.2 46.7 50.0 50.0 46.7 48.2

7 6 5 5 6 7 48.9 48.5 45.6 45.6 48.5 48.9

7 6 6 6 6 7 48.2 46.7 50.0 50.0 46.7 48.2

7 7 7 7 49.5 51.7 51.7 49.5

84 A-35

Fig. A23. DR-3 Reacto FluxeG BO Cort r- sa e Midplane for 93% Enriched Reference Core

O. O O / 1 K 0 CM in \ 0 „ Ü 1 \ fast o «pithermal thermal x

IL. in ö-

• p ö 0.0 12.5 25.0 37.5 50.0 62.5 75.0 87.5 100.0 112.5 125.0 X - AXIS (cm)

Fig. A24. DR-3 Reactor - BOG Flux Ratios at Core Midplane <)>,-/<)> - 0.50 mm Fuel Meat Thickness

_c o. a. o o o CM O O O

X Z>

__ fast 00 ö . epitherma__ l .__ thermal

cp . ö 0.0 12.5 25.0 37.5 50.0 62.5 75.0 87.5 100.0 112.5 125.0 - XAXI S (cm)

85 A-36

A.4.4 Performance Results with MEU(45%) Fuel

With MEU fuel and no changes in the current fuel element geometry, burnup studies were performed using the natural boron concentrations corresponding to the 4.2% and 9.2% rig loads of the HEU core. In each case, searches were performed with the burnup code to compute the uranium density required to match the 23. cycly 5da e excesC lengtEO d s han U corereactivitHE e . th f o y

The results of these studies are shown in Tables A8 and A9. The required uranium densities were compute 1.17e b o t d5 g/cm3 (156. 0235 g element r upe ) wit 4.2a h loag 1.17d %ri an d 2 g/cm3 (155. 235 7g elementr upe ) wit 9.2hà g %ri 235C case EO th u e loadd r contenC witan Fo EO . G 4.2ha d BO loadg an t%e ri th , Pu content in each fuel element are shown in Fig. A25. The average 235u burnup of the fuel elements discharged from each core position is shown in Fig. A26 e averag.Th e discharge burnup averaged oveelementl al re th n si cor46.4%s ewa .

Fast, epithermal, and thermal flux ratios at BOG between the MEU and HEU cases with 4.2% rig load are shown in Fig. A24 for a midplane traverse along the x-axis through the center of the core. The fast and epithermal fluxes with MEU fuel are nearly unchanged or greater than those of the HEU reference. In the core, the peak and the average thermal flux ratios are about 0.95. As expected e thermath , l flux reductio core approximatels th ei n ni y proportional 5.8e tth o % increas core th e n 235i e u loadin(seG BO A25) d et a an gFigs 1 ..A2 heave th peae Ayn th t i k wate r reflector e thermath , l flux rati 0.97s i o . A.4.5 Performance Results with LEU(20%) Fuel

Using the fuel element geometry of the current HEU core, the calculations described in the previous section on MEU fuel were repeated using LEU fuel. An additional case with 4.2% rig load was also run with LEU fuel in which the meat thickness was increased from 0.50 to 0.76 mm and the clad thickness was reduced from 0.51 to 0.38 mm, thus preserving the overall thickness of each fuel tube. The results of these studies are shown in Tables A8 and A9. With the HEU fuel element geometry, the required LEU densities were computed to be 2.81 g/cm3 (165.6 g 235U per element) with a 4.2% rig load and 2.77 g/cra3 (163.5 g 235u per element) with a 9.2% rig load. For the case with a 4.2% rig load, the BOC and EOC 23S contenu uP contenC eacf EO o t d h an tfue l elemen showe ar tFign ni . A27e .Th average 235u burnup of the fuel elements discharged from each core position are show Fign ni . A28 average .Th e discharge burnup averaged ovel elemental r n i s the core was 43.0%. clae Ith fd thicknes reducee b mann i n ys 0.3o sca a t d othe m 8m r light wate heavd an r y water reactors e fueth ,l meat thicknes increasee b n sca o t d 0.76 mm without altering the steady-state thermal-hydraulic safety margins. n thiI s case requiree th , d uranium densit s computeywa e 1.8b o 3t d g/cm3. Since the fuel meat volume was increased by a factor of 1.52, and the 235u loading per element in each LEU case with a 4.2% rig load was about 165 g, the uranium densit nearls i y y inversely proportiona fuee th l o meat l t volume. Thus, the uranium density for clad thicknesses between 0.38 and 0.51 mm can be estimated by interpolation. The 235u and Pu content in each fuel element for the case with 0.7 mea m 64.2d m g loaan tshowe %ri ar d Fign ni . A29 average .Th e 235U burnup averaged ovedischargel al r d element abous si t 43.4% (see Fig. A30).

86 A-37

Fig. A25. DR-3 Reactor - MEU (45%) Fuel, 4.2% Rig Load

BOC and EOC Distribution of 235U and EOC Distribution of Pu Based on Fuel Cycle Length Matching Criterion (0.50 mm Meat Thickness)

EnrichmentU : 45% BOG k 1.0877 U Density: 1.18 g/cm3 effi 1.0433 EOC k' :f Fresh Fuel Loading: 156.0 g 235U Cycle Length: 23.5 Days

235 LJ (BOC) 120.7 118.9 118.9 120.7 235lJ (EOC) 110.2 108.0 108.0 110.2 I>u(EOC) 1.08 1.25 1.25 1.08 235 121.8 124.0 121.5 121.5 124.0 121.8 U Content 111.5 112.5 109.2 109.2 112.5 111.5 Fresh: 4056.0 g 1.06 1.32 1.46 1.46 1.32 1.06 BOC: 3167.8 g EOC: 2874.4 g 121.3 122.6 126.2 126.2 122.6 121.3 110.9 110.7 112.8 112.8 110.7 110.9 1.18 1.40 1.38 1.38 1.40 1.18 121.8 124.0 121.5 121.5 124.0 121.8 Average 235U Burnup 111.5 112.5 109.2 109.2 112.5 111.5 BOC: 21.9% 1.06 1.32 1.46 1.46 1.32 1.06 EOC: 29.1% 120.7 118.9 118.9 120.7 110.2 108.0 108.0 110.2 1.08 1.25 1.25 1.08

Fig. A26. DR-3 Reactor - MEU (45Z) Fuel, 235U Discharge Burnups Elemenf o d aEn tt Lifetimes

Burned 7 7 7 7 nup, % 47.2 49.2 49.2 47.2

7 6 6 6 6 7 45.9 44.2 47.3 47.3 44.2 45.9

7 6 5 5 6 7 46.5 45.9 43.0 43.0 45.9 46.5

7 6 6 6 6 7

45.9 44.2 47.3 47.3 44.2 45.9

7 7 7 7 47.2 49.2 49.2 47.2

87 A-38

Fig. A27. DR-3 Reacto U (20%LE - r) Fuel, 4.2 Loag %Ri d

BOG and EOC Distribution of 235U and EOC Distribution of Pu Based on Fuel Cycle Length Matching Criterion (0.50 mm Meat Thickness)

U Enrichment: 20% BOGk 1.0816 U Density: 2.81 g/ctn3 EOC : 1.0430 Fresh Fuel Loading: 165.g 6 Cycle tength: 23.5 Days

235U (BOG) 131.0 129.4 129.4 131.0 235U (EOC) 120.6 118.6 118.6 120.6 Pu (EOC) 2.43 2.78 2.78 2.43 235U Content 132.0 134.6 132.2 132.2 134.6 132.0 121.9 123.3 120.1 120.1 123.3 121.9 Fresh: 4305.6 g 2.38 2.88 3.18 3.18 2.88 2.38 BOG: 3440.2 g EOC: 3151.0 g 131.7 133.2 136.8 136.8 133.2 131.7 121.4 121.5 123.6 123.6 121.5 121.4 2.62 3.16 2.98 2.98 3.16 2.62 Average 235U Burnup 132.0 134.6 132.2 132.2 134.6 132.0 121.9 123.3 120.1 120.1 123.3 121.9 BOG: 20.1% 2.38 2.88 3.18 3.18 2.88 2.38 EOC: 26.8%

131.0 129.4 129.4 131.0 120.6 118.6 118.6 120.6 2.43 2.78 2.78 2.43

Fig. A28. DR-3 Reactor - LEU (20Z) Fuel (0.50 mm Meat Thickness) Ü Discharg Elemenef o Burnupd En t t Lifetimea s s

Burned 7 7 7 7 235 T rnup% , 44.0 45.7 45.7 44.0

7 6 6 6 6 7 42.8 40.8 43.7 43.7 40.8 42.8

7 6 5 5 6 7 43.2 42.4 39.6 39.6 42.4 43.2

7 6 6 6 6 7

42.8 40.8 43.7 43.7 40.8 42.8

7 7 7 7 44.0 45.7 45.7 44.0

88 A-39

Fig. A29. DR-3 Reactor - LEU (20%) Fuel, 4.2% Rig Load

BOG and EOC Distribution of 23SU and EOC Distribution of Pu Base n Fueo d l Cycle Length Matching Criterion (0.7 Meam m 6t Thickness)

U Enrichment: 20% BOG k 1.0823 3 eff U Density: 1.83 g/cm EOC k »ff 1.0432 Fresh Fuel Loading: 164. 235g 3 U Cycle "Length : 23.5 Days

235 U (BOG) 129.6 128.1 128.1 129.6 235U (EOC) 119.2 117.2 117.2 119.2 (EOCu P ) 2.36 2.69 2.69 2.36 235U Content 130.7 133.2 130.8 130.8 133.2 130.7 120.6 121.9 118.7 118.7 121.9 120.6 Fresh: 4271.8 g 2.30 2.78 3.07 3.07 2.78 2.30 BOG: 3404.7g EOC: 3115.6 g 130.3 131.9 135.4 135.4 131.9 130.2 120.1 120.2 122.3 122.3 120.2 120.1 2.54 2.96 2.88 2.88 2.96 2.54 Average 235U Burnup 130.7 133.2 130.8 130.8 133.2 130.7 120.6 BOG: 20.3% 120.6 121.9 118.7 118.7 121.9 EOC: 27.1% 2.30 2.78 3.07 3.07 2.78 2.30 129.. 6 128.1 128.1 129.6 119.2 117.2 117.2 119.2 2.36 2.69 2.69 2.36

Fig. A30. DR-3 Reacto (20ZU LE r- ) Fuel (0.7 Meam 6m t Thickness) 235U Discharge Burnups at End of Element Lifetimes

Burned 7 7 7 7 •nu% p, 44.3 46.1 46.1 44.3

7 6 6 6 6 7

43.1 41.2 44.0 44.0 41.2 43.1

7 6 5 5 6 7 43.6 42.7 39.9 39.9 42.7 43.6 7 6 6 6 6 7 43.1 41.2 44.0 44.0 41.2 43.1

7 7 7 7 44.3 46.1 46.1 44.3

89 A-40

Fast, epithermal, and thermal flux ratios at BOG between the LEU cases with 4.2% rig load and the HEU case with 4.2% rig load are shown in Figs. A31 and A32 for a midplane traverse along the x-axis. These ratios are nearly identica r fuefo ll meat thicknesse 0.7d 0.5f an so m sinc 6m m 0 m e both cases have 235U loadings of about 165 g per element. The fast and epithermal fluxes in the cores with LEU fuel are nearly unchanged or greater than those of the HEU reference. In the core, the peak thermal «0.625 eV) flux ratios are about 0.87 and the average thermal flux ratios are about 0.88. Again, as expected, the average thermal flux reduction in the core is approximately proportional to the 15% increase in the core 235y loading at BOG (see Figs. heave A27)peake d A2th th an 1yn i st .wate A r reflector e thermath , l flux ratio is 0.95.

A.4.6 Flux Ratios Inside Central Thimbl Fuef eo l Element

DR-e Ith n3 reactor, experimental load oftee sar n locatecentrae th n i dl thimbl fuea f lo e element o demonstratT . e explicitl effece th y enrichmenf o t t reduction on the fluxes inside of a central thimble, calculations were done wit modifiea h modeZ R d l (Fig. reactore A33th e regiof )o Th . n labeled THIMBLE at the core center contains pure 020 and has a radius equal to the radius of the experiment space of a standard fuel element. The CORE region containatoG BO m e densitieth s s calculate burnue th y e pb d ar code d an , homogenized ovecore th re volume midplane Th . e flux centee ratioth f t o ra s the thimble are shown in Table A10 for the fast, epithermal, and thermal fluxes for each of the reduced enrichment cases with 4.2% rig load relative to the HEU case with 4.2% rig load. In each case, the fluxes across the thimble diamete virtualle ar r y constant e fuelsU th LE fas e d .d , th an tan Wit U hME epithermal fluxes are nearly unchanged from those of the HEU reference. The thermal flux ratios are 0.94 and 0.87, respectively, with MEU and LEU fuels. These value nearle ar s y identical wit average hth e flux core ratioth e n si that were computed with the burnup code. That is, the thermal flux reductions inside the central thimble are expected to be approximately proportional to the increase in 235U loading of the fuel elements. A.4.7 Reactivity Worths of Fission Products

Individual reactivity worths of ^Xe, * Sm, and the remaining fission products werHEUe th e ,r calculateU case fo LE MEU C d s EO an ,wit t 4.2a a dh % 13 1! 9 rig loade resultTh . s liste Tabln i dshol eAl w only small differencen i s worths amon foue th gr cores. A.4.8 Reactivity Worths of Natural Boron Poisons

As mentioned above, the natural boron concentrations used to represent the reactivity losses due to rig loads and the cold-to-hot reactivity swing coreC werEO e s U calculate onlHE werd e an ye th ther fo dn fixe subsequenr fo d t calculations with MEU and LEU fuels. Table A12 shows the reactivity worths boroe leakagth d nan f o coreC eEO poisonse th wit r hsfo HEU ,U LE MEU d an , fuels for the 4.2% HEU rig load case. In the MEU and LEU cores, these rigs had reactivity worths of 3.9% and 3.6%, respectively.

90 A-41

Fig. A31. DR-3 Reacto FluG BO x r- Ratio Cort a s e Midplane

Fuem 0.5m - /<| 0, >l Meat Thickness

v- «>

•r v Q. - ^ o O 0 CM L. o O O g —_ _ —~^~~ ...... --•••-•• ..--. .•'"••-...••'

CM /

° f / / CO fast 10 ..... ep it he r mal d~ / s / _ \ . _ _ t ne r mal

• C 1

0 0. 12.5 25.0 37.5 50.0 62.5 75.0 87.5 1000 112.5 1250 X - AXIS (cm)

Fig. A32. DR-3 Reactor - BOG Flux Ratios at Core Midplane

9n/qo - 0.76 mm Fuel Meat Thickness

V •** •*• -C o. V a. r^ L. o 0 O CM t. s~** o o ro O 2,o •v ° __^_ — — — """ ..-••••""""" j§". .•'""'•--..-•" "••--•"' U-

•«• i CD • l 0.0 12 5 25.0 37.5 50.0 62.5 75.0 87.5 100.0 112.5 125,0 - XAXI S (cm)

91 A-42

Figure A33. DR-3 Reacto ModeZ rR Centrar lfo l Irradiation Thimble Flux Calculations

• 35OC . oO* j — 3

o es Q

0) 4J O •H 30.50- CM % Q nJ H 1-01) O 01 rH OJ "I •H

0.0— 1 ! *" - ^ o v o c \ o n e • 00 . . r < H O en m tH

92 A-43

Table A10. DR-3 Reactor - Flux Ratios in Central Irradiation Thimble (Filled with Pure U2Û Standarf )o d Fuel Elemen Cort a t e Mid-plane.

e/(t)93 Fast Epithermal Thermal Enrichment 10.0 MeV-5.53 keV 5.53 keV-0.625 eV 0.625 eV-0.0 eV

45% 0.998 0.991 0.944

20% 0.994 0.980 0.869 (0.50 mm Meat)

20% 0.995 0.982 0.878 (0.7 Meatm 6m )

Table All. Reactivity Worths (<5k/k,% Fissiof )o n Product C CoreEO n si s with HEU, MEU, and LEU Fuels and an HEU Rig Load of 4.2%.

Fission Product LEU LEU Removed HEU MEU (0.5 meatm 0m ) (0.7 meatm 6m )

Xe 3.23 3.26 3.28 3.28

Sm 0.59 0.60 0.61 0.61

Lumped F.P. 2.45 2.32 2.14 2.16

Total 6.27 6.18 6.03 6.05

Table A12. Reactivity Worths (<5k/k,% Naturaf )o l Boron PoisonCoreC EO n si s with HEU, MEU, and LEU Fuels and an HEU Rig Load of 4.2%.

Poison LEU LEU Removed HEU MEU (0.50 mm meat) (0.76 mm meat)

Loadg Ri s 4.15 3.90 3.55 3.58

Tempr Natfo .B Increase 0.72 0.68 0.61 0.62

Extrap. Lengths Restored to Values for 25°C Core 0.73 0.73 0.74 0.74

Total 5.60 5.31 4.90 4.94

93 A-44

A.4.9 Conclusions

The results of this study indicate that it is technically feasible to convert the DR-3 reactor to both MEU and LEU fuels without changing either the core lifetim presene th loadg r eo tri suitablf si e fuel demonstratede sar , available, and licensable. No changes would be required in the thickness of the fuel tubes, and little or no changes would be expected in the. steady-state, thermal-hydraulic safety margins.

e followinTh g table summarize y resultke e th U s LE wit d han HEUU ME , o changefuel n currene f th made i s n ar si e t fuel element geometry l caseAl . s have a cycle length of 23.5 days, the same natural boron concentrations rep- resentin same 4.2e th U coree th g HE d % 4.1e an , Sk/ th g loa% kf ri o d Sk/ k end- of-cycle excess reactivity as the HEU core.

HEU (93% (45%U (20%U ME )LE ) )

Uranium Densit Fuen i y l Meat, g/cra3 0.55 1.18 2.81 235U Loadin Fresr 0 pe g 15 h Element156. ,g 0 165.6 Average 235U Discharge Burnup, % 48.9 46.4 43.0 Average Flux Ratios in Core Fas t- (>5.53 keV1.0) 0 1.01 Epithermal - 0.99 0.99 - Thermal «0.620.9 ) 5eV 5 0.88 Thermal Flux Ratio at Peak in Heavy - 0.97 0.95 Water Reflector With a rig load of 9.2% 6k/k, the uranium densities with MEU and LEU fuels were compute 1.1e b 7o t dg/cm 3 (155. 23J57g elementr Upe 2.7d )an 7 g/cm3 (163. 23S5element)g r Upe , respectively. Normal operating procedures ma ymodifiee b nee o t d operato t d e reactoth e r with thig loari sd since th e end-of-cycle excess reactivit s computeywa -0.9e b o t d%

If the thickness of the clad could be reduced from the current 0.51 mm te thickneso th 0.3fuee , th 8l mm f measo t coul increasee b d d from o 0.5t m 0m 0.7withoum m 6 t changin thicknese th gcurrene th f so t fuel tubes. Thuse ,th fuel meat volume could be increased by a factor of 1.52. With MEU fuel, the required uranium densit r thifo ys cas estimates i e aboue b o t d 0.78 g/cm3 (156. 230g element)r ^upe . Wit U fuel requiree hLE th , d uranium densits wa y calculated to be 1.83 g/cm3 (164.3 g 235U per element). Since the 235U content per element is nearly independent of fuel meat thickness for the same plate thickness, uranium densitie estimatee b y intermediatn an ca sr fo d e fuel meaclad an td thicknesses. Fluburnud xan p performanc l caseal sr fo ewoul e b d approximatel e sam s th thosya e e e tablshowth en ni above.

In the experiment space at the center of each fuel element, the fast, epithermal d thermaan , l flux ratio e approximatelsar coree th e sam -yth s ea average ratios shown in the table above.

94 A-45

JRR-e Th 2 5 ReactoA. r

A.5.1 Design Specifications

The design specifications used in the ANL calculations for the JRR-2 reacto showe ar rTabln ni e A13. These specifications were provide JAERy b d I joina s par a f o tt study JAERd betwee an asseso t IL conversioe nAN th s n potentia JRR-e th 2 f U lo fuelsreacto LE d .an r U froME U fue o mHE t l e currenTh t twenty-four element reference core usin containU gHE s twenty standard MTR-type elements wit curve7 h1 d plate elementr spe d fouan ,r cylindrical-type elements composed of five concentric rings of curved MTR-type plates e cylindrical-typTh . e element more ar se reactive since they containa larger volume fraction of heavy water. At the request of JAERI, the calculations with MEU and LEU fuels were performed with a core composed of twenty-four cylindrical-type elements. Thus 235e ratioe th th ,y f densitieo s s betweee th n presene th U fuel U fued LE e expecte tHE an sd ar l e smallean b U o t dME r than othee th thos rr reactorfo e s studie n thii d s Appendix.

A. 5.2 Calculational Models

A.5.2.1 Cross Sections

Five-group microscopic cross sections for the MTR-type and cylindrical- typ eU referenc HE element e th f eso core were generated wit burnuha f aboupo t d usin EPRI-CELe MW th g0 5 L codee geometrieTh .unie th t f cello s s used for cross section preparatioe th showe A35d r nar an FigsFo n . i n 4 .A3 cylindrical-type elements, all five fueled rings were modeled explicitly. In each case volume th , heavf o e y water associated with each fuel element inside the core was preserved. The heavy water was specified by JAERI to have a light water impurit f abouo y t 3,8% Separate cross sections wit burnua h werd f abouMW po e 0 5 preparet r fo d cylindrical-type elements only with MEU and LEU fuels.

Separate cross sections were also preparereflectore th r fo d s usina g homogenized core source and appropriate thicknesses of heavy water and light water.

5.2.. A 2 Axial Extrapolation Lengths

horizontaA l cross sectioJRR-e th 2f o nreacto shows i r Fign ni . A36. This figure and a vertical cross section were used to model the core in hexagonal-z geometry in order to compute the axial extrapolation lengths for use in subsequent burnup calculations in hexagonal geometry. A planar view of the core mode shows li Fign i n . A37 o controN . l rod experimentar so l loads were modeled. Separate extrapolation lengths were compute eacr fo dh regiof no the reactor with fres corehe e extrapolatio th fueTh .n li n lengths varied between 152 mm and 157 mm over the core. These values are considerably smaller tha DIDne DR-d th thosan O r 3 reactorsfo e , probably because th e impurity level of light water is much higher in the JRR-2. In the heavy water reflector, the extrapolation length was computed to be 366 mm.

95 A-46

Table A13. JRR-2 Reactor - Description of Design Parameters Used in ANL Calculations

Reactor Design Description

Reactor Type Tank Steady-State Power LevelW ,M 10 Number of Standard Fuel Elements 4 Cylindrica( 4 2 , MTR-Type0 2 d lan ) Number of Control Fuel Elements None Irradiation Channels Center Cylindricaf o s l Elements Core Geometry Hexagona3 l Rings Lattice Pitchm ,m 134 Active Core Volume, 1 336 Core Average Volumetric Power Density, kW/l 29.8 Moderator, Coolant D£0 Û Û2 Reflector Burnup Statu Corf o s e Cold Clean Corée

Fuel Element Design Description

Fuel Type Five Concentric MTR-Type Cylinders Curved Plates

Uranium Enrichment,% 93 93 Cylindrical MTR-Type

Plate Thickness, mm 1.27 1.27 Coolant Channel Thicknessm ,m 3.0 2.97 Fuel Meat Material U-A1 Alloy U-A1 Alloy Fuel Meat Dimensionsm ,m Thickness 0.51 0.51 Length 600 600 Width Variable Between 61.0 48.6 and 84.4 Clad Material Aluminum Aluminum Clad Thickness, mm 0.38 0.38 235U Element,g 195 195 235u/Densit Fuen i y l Meat/ ,g 0.64 0.67 Inner 0.21 Outer

96 A-47

Figure A34. Mode JRR-f lo 2 MTR-Type Fuel Element Used for Cross Section Preparation.

EXTRA REGION CU\D UJ ERAT0 0 M O0 R 2 D 87.51 V/0 I 12.49 V/0 A/ O.QE55 0.0.38-»- 0.1485 0.3479

UNIT CELL

Geometry Data Cell Volume Fractions Component Thicknessm ,c Fuel Meat 0.0455 DO 0.8095 0.127 Fuel Plates (17) AI 0.1455 Coolant Channel 0.297 Fuel Meat 0.051 HEU Core Cladding 0.038 235U/Element 195 g 235U Density in Meat 0.668 g/cm3

97 A-48

Figure A35. Model of JRR-2 Cylindrical-Type Fuel Element Used for Cross Section Preparation

Geometry Data Cell Volume Fractions Component Thickness, mm Fuel Meat 0.0330 Fuel Rings*(5) 1.27 0.8726 Inne Rinl rA g 1.27 0.1394 Outer Al Ring 1-00 Coolant Channels 3.00 Fuel Meat 0.51 U CorHE e Clad 0.38 235U/Element 195 g 235U Density in Meat 0.639 g/cm3 *Fuelem dm Lengt0 60 - h

98 A-49

HT-12 TANK HT-11 -lU

HT-10 HT-15

INSTRUMENT TUBE

HT-9

HT-8

. COLUMN DOOR

HT-T

7UB£

HT-6

HT-5 HT-U HT-3

Figure A36. Horizontal Cross Section of JRR-2 Reactor

99 A-50

Figure A37. JRR-2 Reactor - Hexagonal Model Used for Calculation of Extrapolation Distances (3D) and for Burnup Studies (2D) (Dimensions in cm)

100 A-51

A.5.2.3 Mode Burnur fo l p Studies The two-dimensional model used for burnup studies is shown in Fig. JRR-e A37Th normall.s 2i y operated without fuel shufflin fued gan l elements are normally removed from the core when their average burnup approaches 25%. e burnuU referencTh HE e p th mode r e fo lcor e thas agreewa t d upon with JAERI consist liff o f computin so ed en (EOL e th g) excess reactivity correspondino t g 600 MWd of continuous operation. With MEU and LEU fuels, searches were performed to find the uranium density needed to match the EOL excess reactivity U referencoperationf o HE e d oth fMW same 0 eth .e60 cor r fo e

A.5.3 Performance Result HEU(93%r sfo ) Reference Core e resultburnue Th th f pso calculatioU referencHE e th r e fo ncor e ar e summarize Tabln i d frese e th A14ke her f Th .fcorfo e (BOL235u/elementg 5 ;19 ) was computed to be 1.165 ($k/k=14.2%). The corresponding experimental value quoted by JAERI is 15.0% {k/k. After 600 MWd of operation (EOL), the keff was 1.069 ($k/k=6.5%) reactivita r ,o y change burnuo th t 7.7f eo d e %pan du buildup of fission products. The reactivity worth of the fisson products at EOL was calculated to be 2.8% <$k/k. The computed EOL contents of 235u and Pu in each fuel element are shown in Fig. A38. The average 235u burnup in the fuel element with maximum burnup was 21.2%.

peae Th k fast (>0.82 peae 1 th MeVk d thermaan ) l «0.62 fluxe) 5eV n si the flux trap, core reflectod ,an showe r ar Tabl n ni e A15.

A.5.4 Performance Results with MEU(45% LEU(20%d )an ) Fuels

Using the procedure described in Section A.5.2.3, four burnup calculations were performed with MEU and LEU fuels with a core composed of twenty-four cylindrical-type fuel elements. One case with MEU fuel and one with LEU fuel were done with no changes in the present geometry of these elements. With LEU fuel, two additional cases were done in which the fuel meat thickness was increased from its current value of 0.51 mm to 0.76 mm and to 1.0 mm without alterin presene th g t 0.3 clam 8m d thickness e resultTh . f thesso e calculations are also shown in Table A14.

All of the calculations in Table A14 were done using UA1 fuel since neu-

tronics results are not sensitive to the fuel-type employed. X In practice, the uranium densities shown need to be associated with appropriate fuel-types (UAlx, U3Si Ur 23,o Si) . A.5.4.1 Results wit FueU hME l With MEU fuel, the required uranium density was calculated to be 1.44 g/cm3. This uranium density correspond fresr 235a pe o ut shg loadin 8 19 f go cylindrical element, as compared with 195 g 235u per fresh element in the HEU reference core. The EOL contents of 235u and Pu in each fuel element are shown in Fig. A39. The average 235u burnup in the fuel element with maximum burnup was computed to be 19.0%.

Fasd thermaan t l flu xpeake U caseth ratioHE s t a d s san betweeU ME e th n in the flux trap, core, and reflector are shown in Table A15. In all three locations, both fluxes wit U fuee slightlhME lar y greater than those witU hHE fuel.

101 o N)

Table A14.

3 UJRR- Loadin235 2- g with MatcUraniuo t % h 20 Fuem d Enrichmenlan Cycl% 45 e f EnricheLengt% to 93 f ho d Reference Core

Enrichment, % 93 45 20 20 20

Fuel Type U-A1 Alloy UA1X-A1 UA1X-A1 UA1X-A1 UA1X-A1 Fuel Geometry 20 MTR-Type 24 Cylindrical 24 Cylindrical 24 Cylindrical 24 Cylindrical 4 Cylindrical Brunup, MWDb 600.0 600.0 600.0 600.0 600.0 Meat Thickness, mm 0.51 0.51 0.51 0.76 1.00 Clad Thickness, mm 0.38 0.38 0.38 0.38 0.38 235U/ElementC g , 195 198 212 217 225 P25, g/cm3C 0.65 0.69 0.48 0.37 3C PU. g/cm 0.66<* 1.44 3.47 2.38 1.88 2H/235u 979.0 964.2 900.0 862.6 816.9 t"0 BOL keff 1.1651 1,1620 1.1550 1.1530 1.1511

EOL keff 1. 0691 1.0694 1.0689 1.0690 1.0690 235U Burned, g 16.0 15.8 14.7 14.3 13.9 Ave. Discharge 21.2 19.0 17.6 17.2 16.5 Burnup of Element with Maximum Burnup

Note that the 93% enriched reference core has mainly MTR plate-type fuel elements and that the reduced enrichment cores have a cylindrical-type elements. The heavy water volume fractions are different for these two fuel element geometries and the core reactivities and 235U loadings are affected. bBase powea n . do r MW leve0 1 f lo

frese th hr Ufue,fo e P25l ar element ,u P . C235 ^Average cylindricad d an ove R rMT l elements, in fresh fuel element. A-53 Figure A38. JRR-2 HEU (93%) Fuel EOL Distributio DistributioL EO 235f d no U an u P f no Based on Fuel Cycle Length Matching Criterion (0.51 mm meat thickness)

U Enrichment 93% BOL k 1.1651 3 EOL 1.0691 U Density 0.66 g/cm _ eff Fresh Fuel Loading 19 235g 5 U Burnup 600 MWD

235U (EOL) Pu (EOL)

Figure A39. JRR- U (45%2ME ) Fuel EOL Distributio DistributioL EO 235f d no Uan u P f no Base Fuen o d l Cycle Length Matching Criterion (0.5 meam 1m t thickness) U Enrichment 45% BOL k 1.1620 3 1.0694 U Density 1.44 g/cm EOL eff Fresh Fuel Loading 198 g 235U Burnup 600 MWD

103 A-54

Table A15. JRR-2 Flux Ratios

Peak Fast Flux Ratio >0.821 MeV

20% 20% 20% 93% 45% ) (0.5mm 1 ) (0.7mm 6 (1.0 mm)

Flux Trap 1.8617+13 1.007 1.008 1.013 1.022

Core 3.3372+13 1.006 1.006 1.010 1.017

Reflector 1.2261+13 1.008 1.006 1.014 1.027

Peak Thermal Flux Ratio <0.625 eV

20% 20% 20% 93% 45% ) (0.5mm 1 (0.76 mm) (1.0 mm) Flux Trap 1.9427+14 1.001 0.914 0.862 0.827

Core 1.4997+14 1.007 0.895 0.826 0.782

Reflector 1.1586+14 1.011 0.942 0.911 0.894

104 A-55 Figure A40. JRR-2 LEU (20%) Fuel

EOL Distribution of 235U and Pu Based on Fuel Cycle Matching Criterion (0.51 mm meat thickness)

U Enrichment 20% 1.1550 DensitU y 3.47 g/cm3 1.0689 Fresh Fuel Loading 21 2352g U D MW 0 60

Figure A41. JRR-2 LEU (20%) Fuel EOL Distributio Baseu P 235 f Fuen nd o o d U an l Cycle Matching Criterion meam (1.m 0t thickness)

U Enrichment 20% BOL k 1.1511 3 U Density 1.88 g/cm EOL ef f 1.0690 Fresh Fuel Loading 222 5g Burnup : 600 MWD

105 A-56

A.5.4.2 Results wit U FuehLE l Wit U fueunchanged hLE lan d dimension cylindricae th f so l fuel elements, the required uranium densit computes ywa 3.4e b 7o t dg/cm 3 (21 2352g r Upe fresh element). If adequate thermal-hydraulic safety margins can be attained

with fuel meat m (witthicknesse m same 0 hth e 1. 0.3d clam 0.7f an 8m so m d6m thickness), the required uranium densities become 2.38 and 1.88 g/cm,3 respectively. These uranium densities correspond, respectively 235o t , u loadingf o s r frespe hg 5 element 22 u conten P 235e d d Th an .eacun an i t 7 h21 fuel element at EOL for the two LEU cases with 0.51 mm and 1.0 mm thick fuel meat are shown in Figs. A40 and A41. With 0.51 mm thick fuel meat, the average 235jj burnup fuee ith nl element with maximum burnu s 17.6% pwa correspondine Th . g burnup core i th nm thice m wit0 k 1. hfue l mea s 16.5%twa . These burnupe b n ca s increased, if desired, by further increases in the uranium density. Fast (>0.821 MeV) and thermal «0.625) flux ratios between the three LEU e flupeake th xth U casn sHE i t trap ea e th case, d core san d reflecto ,an e ar r show Tabln ni l casese al A15fase n th ,I .t fluxe largee sar r wit U fuehLE l than with HEU fuel by about 1% or more. With 0.51 mm thick fuel meat, the peak thermal fluxes in the flux trap, core, and reflector are reduced by about , respectively , 10%6% 9% d ,an . These thermal fluxe furthee sar r reduces a d the thicknes fuee th l f measo increaseds ti .

A 5.5 Conclusions The results of this study indicate that it is technically feasible to convert the JRR-2 reactor to both MEU and LEU fuel using the same cylindrical fuel element geometry tha useds i tcurrene partn th i , U n cori ,tHE e without changing either the core lifetime or the present rig load if suitable fuels are demonstrated, available, and licensable. With MEU fuel and no changes in the geometry of the current HEU cylindrical elements, the required uranium density was calculated to be 1.44 g/cm3. This uranium density corresponds to a 235u loading of 198 g per element, as compared with 195 g 235y per element in the HEU reference core. The average burnup in the fuel element with maximum burnup was reduced from e flu th 19%o x t abou n I trap.% 21 t , core d reflectoran , peae th ,k fasd an t the peak thermal fluxe slightle sar y greater wit U fuehME l than wit U fuelhHE .

Wit U fued unchangehLE lan d dimension currene th f so t cylindrical fuel elements requiree ,th d uranium densit s compute ywa 3.4 e b o 7t d g/cm3 (212g 235 element)r pe u adequatf I . e thermal-hydraulic safety margine b n sca

maintained with fuel requiree meath t, thicknessemm d 0 1. d 0.7an f s o m 6m

3 uranium densities become 2.38 g/cm 3 (217 g 235{j per element) and 1.88 g/cm (225 g 235u per element), respectively. The average burnup at EOC in the fuel element with maximum burnup in the LEU cases with fuel meat thicknesses of 0.51 mm, 0.76 mm, and 1.0 mm were computed to be 17.6%, 17.2%, and 16.5%, respectively compares ,a d witU coreHE h e .21.2 th Thesn %i e burnups coule db increased desiredf i , , with further increase uraniue th n si m density.

e fasTh t O0.821 MeV) flufluxee th x n si trap , core reflectod an , e rar larger with LEU fuel (for fuel meat thicknesses between 0.51 and 1.0 mm) than wit U fue hHE moreabour y lo b % 1 .t With 0.5 m thic1m k fuel meat peae th ,k thermal «0.62 flue fluxe) 5th eV x n trapsi , core reflectod an , reducee rar d by about 9%, 10%, and 6%, respectively. These thermal fluxes are further reduced as the thickness of the fuel meat is increased.

106 APPENDIX B

Unclassified

Paper to be presented to the Consultants' Meeting on Heavy Water Research Reactor Core Conversion to Reduced Enrichment Uranium, Vienna, Austria 6-7 May 1982.

Enrichment Reduction Calculation e DIDth Or Reactofo s r (Extended Calculation e Consequenceth n o s f Reductioo s e th n i n DIDO Reacto Enrichmenr% Fue20 d l an Suppliet% Levels.45 o t s )

G. Constantine M. Javadi Elizabeth Thick Materials Physics Division, AERE, Harwell, Didcot, United Kingdom

107 On the Consequences of Reducing the Enrichment of the Fuel Supplies for the Harwell Materials Testing Reactors

INTRODUCTION

e possibilitTh bees ha yn raised that DIDO/PLUTO type heavy water moderated

reactor operatee b n sca d with fue f loweo l % enrichmenr 75 thae th n t material

currently in use with the object of increasing the proliferation resistance of

the fuel cycle. This paper sets out to examine the reactor physics aspects of

enrichment reductions to 45% and 20% for Harwell's MTR's as part of an IAEA collaborative exercise currently being conducted to examine the topic in a more whole th er genera clasfo y heavf o swa l y water moderated reactors.

The DIDO Reactor

DIDPLUTd an OHarwelt Oa y cycleda l 8 operat,2 a consistin n eo day4 2 sf o g

at full power (25.5 Mw), followed by 4 days shutdown for maintenance work and

d fuerian gl changes. Dependin reactivitg g ri upo e nth cycle yth o loat er fo d

come, the number and disposition of fuel elements changed is arranged so that

the cycle will finish with approximately 1.5% ôk/k excess reactivity remaining.

This is a compromise - more available reactivity is costly in terms of fuel

element consumption while with less the reactor's ability to restart after

an unscheduled shutdown before poisioning out mucs i , h impaired. Wite th h

current reactivity loading approximately 1/3 of the core is replaced at each

shutdown. It was decided at the outset of this study that the use of lower

enrichment fuel shoul e consideredb sam e basig loae th termn eri th i d f so n do s

of total absorber cross section together with the criteria of 1/3 core replace-

ment every cycle and an end of cycle (EOC) available reactivity of 1.5% ôk/k.

Fuel shufflin normallt no s i g y practise DIDn r operationai d fo O d an l

safety reasons. The average sojourn of fuel elements in the reactor is 3 cycles,

those neacore th re centre averaging lesd thosan s e neaedge th re more than

3 cycles. Examination of the records reveals that over many cycles the mean

burn-up at EOC of fuel at all positions across the core is sensibly constant

108 at -32%, those central positions whence fue s dischargei l d more frequently

compensating by a higher burn-up rate. A useful theoretical model of the reactor

under average EOC conditions has been developed in which the representative

fuel is based on a mean of 3 fuel elements which have been exposed to 1,2 and

3 reactor cycle totat sa l fission energy release rating f 1.167o s , 1.06d 3an

0.94 w respectively9M o efforN . s thui t s mad modeo t e l fuel elements individually

ende ath f theitso r first, secon r thiro d d reactorcyclee th n i s , although

the averag e modee closa elemen th s i ln e i approximatiote us tha e tw n a o t n secons it f do cycled en elemene .th t a t

A notional reactivity balance table can be drawn up for both beginning and

end of cycle (BOC and EOC) as shown in Table 1. The allowance for rigs includes

4.5% ok/k absorbe y in-corb d e reflectoe rigth sy b wit % r2 h rigs. Currently

gadoliniu borod an me employe nar burnabls a d e poisons formee ,th a shorr t term 135 compensato t e on e cycle e earlr build-uth X e latteefo n th i ,y matco f t ro p h

the .reactivity loss from fuel depletion. It is inefficient in leaving 1% ôk/k

absorption at EOC. Although the majority of this remains in the poisoned elements

that have completed their first cycle wits a , h representativrige th s e absorber

used in the fuel elements is distributed evenly across the core in the theoretical model. Appendi give1 x s detailreactoe th d fuef so ran l element design.

Table 1

Reactivity Balance for DID Beginnint a O g f Cyclo d aneen d

Absorbers BOC EOC

Rigs 6.5 6.5 Longterm fission product 2.3 poisons 1.15 Burnable Poisons 5.0 1.0 Xe -i- Sm 1.3 3.9 Temperature* 1.4 1.4 Control Absorbers 4.7 1.5

20.05 16.60 Fuel Depletion from 3.45 6.90 Clean Core

23.50 23.50 •Los f reactivitso y frocole th md conditio 20°t a n C uniform temperature th o t e full power working temperature distribution.

109 CALCULATION METHODS

The reactor physics tool used at Harwell is WIMSE , the Winfrith ^mproved

Multigroup Scheme suit,a f linkeeo d reactor physics codes whic s beeha hn used

extensivel r lighfo y t water, heavy wate d graphitan r e moderated thermal reactors. e variouTh s code n thii s s suite intercommunicat seriea a ef vi interfaceo s s

carrying intermediate results suc neutros a h n spectra, number densities, cross

section data, geometry specifications etc., written on disc and read as required

in subsequent stages. Such functions as lattice cell calculations using

collision probability methods, spectrum weighting and cross section averaging-,

group condensation, editing of spectra to produce reaction rates, collision probabilit d neutroan y n transport solution dimensionae on n i s l geometry,

diffusion calculations in two and three dimensions, burnup calculations etc.

can all be performed consecutively by simple commands.

The calculatione coursth f WIMSe eo th Ed modulesan s involve n thii d s

study are described briefly below:-

) WBUR(i N take initiae th s l fuel element specification, carriea t sou

lattice cell calculation in 69 groups using module WPERS as a "subcontractor" and computes reaction rates in the fuelled regions. The

calculated spectra are assumed constant and fission product generation

and fissile species burn-up determined over a user-specified time

interva bead an lt rating. Thi dons i s e stepwise over small subdivisions

of the time interval. After this the burned up fuel and fission

product concentration consideree sar fresa s a dh lattice cell

specification and the process is repeated by computing neutron spectra again.

(ii) Haying computed fuel and fission product concentrations at EOC1,

• EOC2, and EOC3 for the 3 cycles a mean is taken for insertion in the

reactor model as the representative fuel. A lattice cell calculation

is performed by WPERS on this specification, rigs and residual burnable

110 poison being represented by boron dispersed in the central 50 ma

diameter facility.

(iii) Module WSMER computes mean fluxes across the cell and obtains weighted

mean r eacfo sh isotope.

(iv) WMER concernes i G d with data assembl core d reflectoth an e r fo y s a r

a precursor to WSNAP the diffusion code module which first carries

out a one-dimensional reactor calculation (R - °°) in 69 groups.

Usin neutroe th g n spectra radial distribution t thesi n collapsee th s

diffusion data to 7 groups for 4 distinct radial regions in the core

reflectore th n i an 9 d , encompassing D_0, graphit d reactoan e r tank

regions e reflectoTh . r regions include concentration materiale th f so s

representativn i e rigs.

(v) WSNAP is then employed again in its 2-dimensional version to carry

out an RZ computation in 7 groups using the diffusion constants

obtained froR-°e th m° calculation.

At each stage, "trace" material specifiee b n 9 grouca s6 o thae s dp th t cross

sections obtained fro librare collapsee b th m appropriate n th ca y n o d grou7 e p

spectr computinr fo a g reaction rate distribution grou7 e modelZ th pR n i s.

Fig. 2 shows the geometrical configuration of the R-Z calculation.

Calculations performed

The basic existing reactor specification with 75% enriched fuel at 205 g 235 U per element serves as the reference point for this study. Two alternative

% hav20 eenrichmentd beean n% considere45 f so calculationd an d s performed with

differen r fuepe l t U elemenmasse f eact so a t h enrichment235 , makin case5 g s

in all as listed in Table 2a. Table 2a also shows the intermediate and final

result multiplication-factorf so chaie th f calculation o nn i s infinits- e

) reactoZ lattic- R ( r modelsd e an cel ) l°° . followe- R ( y b d

Ill Tabla 2 e

k235r f k f r Enrichmen- * t Wt ef° Uf eff ° g/fuel element lattice cell R-=? cale. R-Z cale

75 205 1.4810 1.2070 1.0216

45 210 1.4634 1.2020 1.0234

45 220 1.4738 1.2137 1.0352

20 220 1.4191 1.1786 1.0121

20 240 1.4396 1.2030 1.0376

Tablb e2

45 208.5 1.4617 1.2005 1.0215

20 227.4 1.4273 1.1885 1.0222

The results for the R-Z calculations are plotted in Fig. I, from which

% enrichmentestimates ii t20 d an d% fuee 45 ,thath l t a tweight s necessar. o t y

achieve the same EOC k-eff as the reference 75% case are 208.5 and 227.4 g 235 U respectively. These weights were incorporated into fuel element specifi-

n througcationru samd e an th she burn-ue procedurth f o p d estageen froe th m,

the number densities of all the fuel and fission product isotopes being derived

as weighted averages from the final output of the WBURN stage of the appropriate

finae caseTh Tabln l si . k-ef e2a enrichmento tw f e valueth r sfo s usine gth

estimated fuel weights appear in Table 2b and ars in satisfactory agreement

witreference th h e case.

Flux* distributions have been plotted alon radiua g s froe centrth m f eo

the group3 corsame n th i ee Fign n . i si grouThes3 . e pear structurs ea

tha benchmare t th use n i d k calculationsMeV0 1 .- 0.62V - 5.5- ke 0 V 35 e z ,vi reference Thth er result% fo enrichmen e75 e ar Figcasen g si 3 n .5 I .20 t

orde shoo change t r th w performancn ei reducinn eo enrichmene th g readila n i t y

apparent manner, the flux levels in the 45%, 208.5 g and 20%, 227.4 g reactor

are showplotte5 % case Figs'd n 75 ni an e ratio s . a d.th 4 o t s

112 Conclusions

It can be seen that apart from the problems of accommodating the additional

uranium in the fuel element design .currently being addressed in another part

of this stud e consequenceth y reducinf so enrichmene t th gno fuee e th ar l f to

too serious from the reactor physics standpoint. This is true particularly

lessee casth e if th nreo % levereductio45 e l th wher o fuee nt th el content

elemente oth f maintaio st mus% raisee ~2 tb n y currenb d t operating conditions

in terms of rig loading, fuel element consumption and EOC available reactivity. % enrichmenEve20 % increast na 11 e tth fuen i e l loading wors e leadth o t t a s

a 17% fall in thermal flux at the core centre and much lesser consequences

elsewher ehighee whilth n ri e energy group flue th sx change e rathear s r less

significant everywhere.

The fall in reactivity worth of control absorbers in the core will follow w approximatella M consequencee 1/ th a d an y f thiso s wil le lookeb nee o t t da d

seriously. Howeven probablca t i re argue b y d thay evenan t t initiatina g

reactivity accident will similarly be reduced in magnitude by the same factor so

the actual reactivity excursion, assuming it is terminated by absorber

movement on the same timescale, will actually be less. Although the calculations

have not been carried out for DIDO itself, the results of the benchmark calculations indicate that althoug prompe th h t neutron lifetime will fall

slightly, the response of the reactor to very fast transients involving the 238 fuel temperature, is considerably improved by the presence of more U in the

core via the Doppler effect. Overall temperature and void coefficients are

not change y significandb t amount resula s a sreducinf to e enrichmentth g .

References

1. Askew, J.R., Payers, F.J., Kemshell, P.B. A general description of the

lattice code WIMS, JBNES Vol. 5, No. 4, pp 564 (1966).

113 Appendi1 x

Reactor Specification

O cooleD a moderateDIDd s an di O d enriched uranium thermal reactor,

operatine corTh e contain. 25.t Mw a g5 5 fue 2 slatticm m l 2 element15 e a n o s

pitch in a 4-6-5-6-4 configuration, the centre row being staggered by J pitch

from an otherwise square array. The DO tank made of aluminium is 1851 mm internal

diameter surrounded by a 458 nun thick radial reflector of graphite. The tank

has a dished bottom with a similar graphite reflector underneath.

e specificatioTh fuee th l below t f elementno ou t .se currenn s i i e tus

They comprise 4 concentric co-extruded fuel alloy tubes with an unfuelled wrapper

tube e fuelleTh . d lengt 609..s i h m althoug6m h becaus e facth t f eo tha e th t

uranium conten s taperee effectivende i t th th s , t a ovedmm lase th 2 ert6 length

. mm 7 i54 s

Uranium Enrichment,% 75 Outer Tube m O.D.m , 103.0 Plate Thickness, mm 1.68 Coolant Channel Thickness, mm 3.09 Fuel Meat Material U-A1 Alloy Average Fuel Meat Thickness, mm 0.732 30.564 Fuel Meat Dimensionsm ,m 31.29= R Tub , 6R : e1 o i 36.06= TubR ^ 6R : e2 35.334 o i 40.83= TubR , 6R : e3 o i 40.104 45.60= TubR ^ 6R : a4 44.874 o i Clad Material Aluminium Average Clad Thicknessm ,m 0.474 Average 23 5U Density in Fuel Meat, 0.480 g/cm3 235 U/Standard Fuel Element, g 205 235 Averag DischargU e e Burnup,% 40-50

114 Flg. 1 k-effective vs % fue20 l d loadinan % 45 g t a enrichment rTT ii n-Ti '«-pi TiTTpin üf i i il f"[' ' IJiW™ ^liTTprriii ii! •,i j.i MI !ihi::ii'ii::|:i;.'ii;!i|: ^'ihi;! I'!:'" : :! '' •! ^lülllliü Iiiii i , i .!i! i i !: .i Üil jü jli... iiSivii» Ii ji 18 l! iii'iiii $ j, jl: "l Üih'ilil ' l II! ni ji;! iii iii !: ! |! 11 l l ii. :. „, |, ! ii! | !||i !'h iil-i! I i » iiii! til | : ::1! i i ,|l 1 : . -t T i 'i •-i 'i'llli iiii ' ' •• 'iii H,i ii il'" I i in ii : ,î ' I I 1 i' ! ii i i' 'si;ni In •!ii , i | ; 'i'lN'l I 1lift! -« ! ' 1 !i| Ü! l! 1:1 i ! Ill ill! L l II. ü! '" i|!ü«iW..|- üüllLEllj 17il^ ilülî li .! - iüiüp•: tj. jiS ;j|!!!•: !i .i!'• ...illI îjiîiiiiiiliiii i li! l! ! |!; lü! jjlj Iü J!:ii:!;iill!!i! iiii1 i iüMlilüli! ui, Ji! Üi !'! l ! ii! il l-!' •'> IÜ, 4ii UPF ,ii 1,1 • l. . ' • L' .ill UM ['i. If l ] iiiIt i •;.} iji,-j:i 11 : : ! 1 6 iü: •:;• . i t n i i Ln Ci un , , 5 !i i' . ii i^ ! " hinan "' l1' !•:• ,; , ; '[|i.ili - i ![ jjil!il.,il|JJii|l:HU;{i::i'':,aN|| üjiiiitjjji" :;:i i ii:!i jü : i Lj i 'i!l ÜL i|!i N•»•ii'lili'-• HKi : : î M;i|:ju;ii ii!î!i iHii'iii! "l/llü n •iii'.'i'i.Hiiiï i i üü ijiiÜII .l'. , i ' l I ijl i : I . i r 'i , : ! 'j | !h i ; | r i! ,-!! .»! ' ij i;ii & §. ,;i • ! '; i j. :i. si :ih ii i\ : lui !üil . i : 11. il.'i 'i:: i i ' ''', ij'i ';M Mji :,]' jii i| ÜlÜiÜlljj; i ' ' !'i• | ••''l!i!S!;Î!!i|!!ii|!li! !i!:i l!!ii!!'i' iiii .n•ini .1" . iijt ,l i... i il !it. >, il i . ii l, ! :|i . M |.|niI .. * i , i 'i.M 'I i iiii .IM ti i«'"" ' t'!!;!!! iiii 1 Ü! iHiijiiiililif liiliii! «!'[!$ ^l 4i, "' h'jNuii 111 1 iiii 'Ü' . . i 11 -it| f». ... .1*1 4*«* i-j*- »« ;*•"* •*•'•* •;' »î " '*• li'iK ! !0! ,: ,.l ;!| II1 l | Jli hv'Hhl jj|! |i;r i. i 'riiii.M .'•' : ! ! : l ii!.,i. il , :t ' ; !: 'jl! ;:!: jl ! ! l ! Jlj! l ! i:i, | i i j ; ;; ! l h iii n 'i i i' ! 'Ii l il i i i l ,' u i!'i ;. n 'M ii ii 'llj ill! :i|: l ,:!! !!•;• Üi i .!.| ! . i i| ,!jj . fJ! ! IÜ l al ! ! i's! :" !: ! 4! J ilü !|ii :' l i:!i :., .4.; :..' .1.! ]! i ... i ;!: 'hiiel i r t* ,miep i_ i ri e i .,,. Ü'rv , _1 , ,__i.i/_ , ' i '"'*• i ! ••' / ' ' ,ii. ji|jii:; j i;ili'!, in '•'''! '• '" • |!}i i'il ,K 1 t •'.:»•: li hui !- i 'i• h ü • •.!u •' -l ' , : îi! iiil iiii : « .:i :!i^ ;ii! > "'ii. |i|h:!::)!ipi.vhuKiüi U : !ü i Jl I I •}•. ; M| ,j. i.i! i l, jjlj[!| "«'M i'|i Iiii iiü i'! •rt : i j iii»| ! i !:ii i i i 'ji:i i ijlf.ij i i:] i ij ! ' iir n.: in. V _i i ù V,!! i|!'|j',| n^.liij!;?!!'!': ! ll!1 !l 1l i'It V i! • !•! ;: HI' ;. i .!;, /!' i' . i •'».ü U i ill •> It *]i .'!' l if I ''•ii;-: i'ii.i' i !••• " « j.Ji: •l!!ii!iEI;;M!iÜ:k;,;i liuuiiîi ii!i,:i!iîl Üii !L. :lj ii i Jiiüi .l\: .!U iiü |.' :.| ,J . i i' 1t 1 |; 1 1 , ,,;;•; ii u.iaib'iii^fciilliJiiiiliiii'iiiiii iiin'li ir jtJii i!' 'i : : i iii! T r \ 'h ji-!:! '!!! ^!.!.! . •:n •' :. i.i . : . ! , '.': i l j i • ' • • i H l ! 1 . r f ' i|i:;: h. t.h ^. l.!;..u -.i,:i::l ! , ' .' " ! . •' !:! l ':i li: i'!, ü-t f\ , M. n, '"; •;" Mil if - 'l, : i ! 15 1 : üi •'•' :i/ , ''J ' ' •t. i i: ' i , !|!' -! l j. : : n 'ih •i, •:;'•" ' «'ü',;j, i iîlJi'l • i i! i j ujj||'äf"d ^ü .i' i ii. ..!! . I J :._: :LL M J^. !:jJ .:... iu jiU 'iii üii ilii:!: ^!]iijiiy.;{ ..J. l l: -|l l!i. '., i , IJ i l • il.l lui . i'i •i tr j: iüi i'i : ; r !...!''' "i ii 'i'M « * ni i1' i i'!"!1 . • ! !:,'j':!i iill luliil l • l •') n!; ,i «'l- '" ..u/:.! i i'! i, 1 ! MM '!• i '•!• 'i!' '• i i • 'i ii U " iiiÜJhiÜi'iil.u'-ir :i !: r-!M ; J >. :.!: .li. j! H 11 •i .... i, .,; ri ',. SiKiiiliii'i!; !i ^ i 'il i i l :• .'i. :• ,:!'!•! IL- .«.! ., r ''!! M l'l |i,' ii i i ' n . r. .ii1 •i, l,- ,t, , i- | 1 ,1^ ' ••!- ! i'i» & ' :J U . .l ' sii.. i 'I : a ! t :J. Jli! ..li . • V.-iülX !:!, i i' " ' h l ' •~i •'! i ' ' ] 1 !•!' > • i "in ' i . i. ii M» .h J.I il.1 !l' . i !i j . Pjijjil ,;,!!, < ! . 1 . . i . .i|. , , . . i|| i . i,, - |. u i ( « |, l il ,i i' ;.:;;:••:'i; i;i!i- i •_ .'.. .1 220 240 235 wt U per fuel element axis of rotation V

D2° * core reflector graphite rigs

Al tank

Fig. 2: Representation of DIDO in R-Z calculation

116 Fig: Radia3 . l Flux Profil Cort ea e Centre Level % Enrichmen75 s t MeaC EO n

-~:~~~"- forpt» '•'—-"'"

H^t:—TirEn-dH—-—. =~:Oroup:':H:=Energy : Range •:-.:• :-'-

" • ~ " "

x •3

-=l=

50 100 150 radiits from core centre (cm)

117 Fig. 4: DIDO Reactor Core Midplane EOC Flux Ratios c(20)/$(75)

1.00

.96

ca K .92 ,., X

.88

.84

. ._...>...... _...... i.»' Tir- T" " * ° *"~' t : . • —"

^=7-— •Tr-l-'-^frL7:-r:-: - —— ! ~.tr.--"™:::= •n-r-'r-J'-Srhsr...::.— —.:•".-— i£l— — ^--rJH.'}:".

--.I fe-n-2-"-:- "J-~ ---•Hré^—ïT r^F—lT:-:.".— --:t--:-— --^:- •r.-:r:~ ::=::-:T_„r.:::-;.=~_t-:_._:.1::-:::~l ... -"j .- :: 5: DIDO Beactor VdJ. C Midplane ÜÜG Ratios i'ig. (45)/*(75) .^tp^;;::»-.^-.- ;i:u :.- ]

•Jas t

:r. i: therma l.ufUUiHt^Jz~ .-Jr~i i L

1.00

.96

.92

-T1——

"" "n " -i" "rt~-7T':zr::^:i^ Z^TT: in

118 i APPENDIXC

ENRICHMENT REDUCTION CALCULATIONS FOR THE HIFAR REACTOR (AAEC, AUSTRALIA)

This Appendix primarily presents two self-contained study papers : Paper Cl (Neutronics) summarises the results of a recent update and expansion

of basic survey calculations carried out in 1979, which were among the first to

show that fuel cycle parameter maintainee b n sca reducet da d enrichment with much

smaller increases in U loadings than previous investigations, mostly based on 235 "clean cold k ... or k " matching, had led reactor users generally to believe, et r o» The mode methodd lan s used differ little froe formth m develope 1979n i d d ,an are essentiall same thosys th ea benchmare eth user fo d k calculation (Section 2.4). Further demonstrates i t ,i d that results woula little f db o e eus changee th y db

more complex "core-follow" model. Fofirse rth t enrichment comparison assumes i t ,i duraniuo thaw/ 2 t4 m still

represent uppee sth r limi f "currento t commercially available proven U-Al fuel

technology therefors i d " an maximue eth m fuel meat loading presen t whica n hca t be considered as practicable in the near to medium term. On this basis, it is

shown that the fuel cycle and burn-up parameters of the high enrichment mode of operatio quite b n en ca readil y maintaine intermediatt da e enrichment {notionally

45% 23SU) levels, without necessitating any change to the existing (HEU) fuel element geometry. At the 20% 235U level, however, thicker plates, and consequently smaller coolant flow annuli woul necessare db accommodato yt uraniue eth m loading required to.maintain the same conditions. A typical example of this option is examined. Further calculation thee sar n presente 235% 20 U r enrichmenfo d whicn ti t hi is assumed that UaOs-Al fuels wil e successfulllb y develope proved an duraniur fo n m densities enabling the necessary 235U loadings to be obtained without change of

fuel element geometry. Two examples are considered.

Finally, a brief examination is made of the consequences at 20% 23Su enrichment compared with standar reactoe fuelU th dHE f ,i r wer. e MW uprate 5 1 o dt

Pape present2 rC resulte sth coolanf so t flow stability calculations, carried

ouparallen i t l wit neutronice hth s calculationobjective Th o t . s Cl ewa f so

119 quantif reductioe y th thermohydrauli y ke n ni c safety margin coolanf si t channel

widths were reduced to accommodate the thicker fuel plates necessary for enrich-

ment reductio 235% 20 U weigh2 o 4 undent cenr e trpe th t Uranium limitation. awart no similaf e o ar e rW studies having been published elsewheree th d ,an

results may be a valuable guide to reactor users contemplating a fuel element

geometry change in conjunction with enrichment reduction. It must be stressed,

however, that the results, although believed to be reliable, are based on general

data correlations only hav d beet ,an e no n verifie direcy db t experimental measurements. For these and other less definable reasons, it is considered that reductions

in thermohydraulic safety margins, even when relatively small, would be likely to

encounter acceptance difficulties with most licencing authorities.

Conclusions

The overall acceptability to a Research Reactor Operator of any proposed

reduced enrichment fuel element wil influencee lb many db y factors which have

receive consideratioo n d thin ni s Appendix. Important example sucf so h factors

would include the incremental costs and demonstrated reliability and irradiation

performanc elementf eo snewly-developee baseth n do d high uranium density materials.

stresses Ii t d tha conclusione tth s which follow have take accouno nn sucf to h

factors and their acceptability, but are based solely on the neutronic/thermohydraulic performance features addresse. C2 paperd n di an l sC (1) Principal fuel cycle/burn-up parameters of HEU fuel can be retained

at either intermediate (45% 235U) or low (20% 235U) enrichment levels at quite modest increases in the 235U content per element.

Attendant change reactivitn si y coefficient similad san r safety-

related parameter smale sar eithet la r enrichment.

(2) At 45% 23SU, fluxes and operational flexibility suffer relatively small degradation, and would probably be acceptable to most operators. At 20% 23SU, the penalties are much more severe, particularly in relation to in-core thermal flux. This situation

will be virtually unaffected by development of improved fuels, and is likely to be of grave concern to most operators, both in

120 absolute consequences, and in effect on relative economic competitive-

othes vi ra nesreactos svi r types.

(3) Wit provisoe hth s above, reductio 23S% 45 U o enrichment n feasibls ti e

in principle within current fuel technology constraints. Reduction to

235% 20 U enrichment woul thesn di e circumstances require thicker fuel

plates with attendant reductio thermohydraulin ni c safety margins, which

may not find licencing acceptance. Changes in the self-limiting transient

characteristics for a reactor so changed have not been examined, but are

not expected to cause serious difficulties. It is nevertheless essential

that this be confirmed by appropriate studies before an inevitable

commitment to a particular enrichment reduction course is made, and it

is intende staro dt t thi sneae th wor rn i kfuture .

Althoug) (4 maie hth n sequenc enrichmenf eo t comparison calculationr fo s i s

reactoW M 0 1 r power doe t ,t appeai sno r tha conclusione tth s woule db

change significany reactoe an th n f i di r y wertwa . e MH uprate 5 1 o t d

Still higher powers, for which substantially higher 235U fuel element

loading normalle sar y used, would require further appraisal.

121 APPENDIX C - PAPER Cl

NEOTRONIC CALCULATIONS FOR REDUCED ENRICHMENT FUEL IN HIFAR

G.S. ROBINSON and B.V. HARRINGTON Nuclear Technology Division Australian Atomic Energy Commission

0 Descriptio1. HIFAf no R CorFued ean l Elements HIFADida s oRi class reactor wit fue5 h2 l element 152.a n s o squar 4m m e

pitch in a 4,6,5,6,4 array with the central row displaced one half-pitch. The

heav moderatoye wateth r reflectord fo ran containes ri cylindricaa n di l alum-

inium tank, 2 m diameter and 12.7 mm thick. The bottom of the tank (16 mm thick)

is dished, but in RZ models of the reactor, is assumed to be flat, providing a

bottom heavy water reflector uniforml reflectoyO Da 0.4 p 5thickto m re Th .

extends 0.7 8abovm active eth e core. Further radia bottod lan m reflectorm 6 0. f so thick graphite are located outside the tank.

A large number of horizontal beam tubes and vertical facilities extend into

the heavy water reflector excepl Al .smalfacilitN e tTA th l2 y have been represented

in the calculations. The 6 "signal arm" absorbers which control the reactor have,

however, not been included. The reactor operates at 10 MW on a 28 day cycle (about

24 days at power). The standard fuel element consist concentri4 f so c fuel tubes, eac 0.6f ho m 6m

thick U-Al alloy 'meat' clad in 0.43 mm thick aluminium. The active length is 603 mm. Each fuel tube is made up of 3 curved plates welded together and the

consequent aluminium seams between plates reduce the notional volume available for

fuel meat by 9.1%. The inner radii of the fuel tubes are 30.39, 35.29, 40.19 and

45.0 . Inne9 mm oute d ran r aluminium tube insidf so e radii 25.3 d 49.9d an 6an m 0m

thicknesses 1.6 1.5d 3an respectivelm 9m y complet elemene eth providd tan coolan5 e t

channel widtf so h (inne outero rt ) 3.4, 3.38, 3.38, standare 3.3 Th 3.2d 8. an 9mm d

element contains 150 g 235U at an enrichment of 80% 23SU. Five fuel element types were considered in this study as detailed in Table 1. The above fuel element geometry (including 3 plate construction method) was

123 maintaine casesl al n ,i d excep r fuefo tl typ wherA e20 thickeea r fuel meas twa

required (see below). In this case the mean fuel tube radii were maintained and

the coolant channel widths reduced to 3.23, 3.04, 3.04, 3.04 and 3.12 mm (inner to outer).

standare Fueth ls fuei U typ dHE A luraniu% meate 80 witwt th 7 h1 ;n m i

Fuel type 45A is a 45% 235U enriched element using UA1 -Al of 29.9 wt% A uranium;

Fuel type 20A, als-Aljrepresentl oUA situatioe 235% sth 20 U r enrichmennfo t

limie a tth currentlf to y proven fuel technolog uraniu% wt eithe2 r (4 fo my r

UAl or UsOg-Al). An increase in fuel meat thickness was required to achieve

the necessary 2 o cU loading;

235 235 Fuel type B (17s20 g 0 U/element C (1620 g 0 U/element)d )an Ue 3,Oar 8-A1,

the probable favoured long-term fuel materialwithout restrictio uraniun no m > density. The cases correspond to 67.1 and 64.9 wt% UsOs respectively.

2.0 Reactor Models methode Th s use thesn di e calculations wer same ethosth s e a e th use n i d

benchmark calculations (Appendix F3) where they are described briefly. The

generatio crosf no s section latticn si e burnup calculations differs slightly from

the benchmark beiny ,b g performe constant da t flux rather than constant power.

HIFAThY eX- R model used include detailesa d representatio botf no h horizontal

and vertical facilitie heave th yn s i wate r reflecto typicaa d loadingrg an lri .

It has been found necessary to include this detail in order to achieve criticality calculatione inth obtaid san n reasonable flux distributions. Although such detail

is not essential for the present purpose, in comparing different fuel enrichments

for reactors with large amounts of reactivity invested in rig loading and fuel

burn-up, it is preferable whenever possible, that the rig burden be included

directly in the calculation. The horizontal beam tubes were the most difficult to represent, and were

treate approximatn a y db e model, wit resulte hth s S normalisecalculatio Z R n a o t d n majoe obeafH th r10 m tube reactivite Th . y radiae wortth f hlo reflector facilities

include 5.0%s additionan i A thi n di .y swa l reactivity wort 2.2f ho % (measured

124 value) for reflector rigs was included by inserting the required amount of thermal

g locationsri absorbe e th typica t A a r. l in-cor g loadineri f 2.3go % spread over

5 fuel element alss owa s simulate insertioy db f localiseno d thermal absorbers.

Bücklings for the XY geometry calculations were obtained from RZ calculations

which included smeared representation e structurth e radiaf o sth axia d n i ean l l

heavy water reflectors. Centre-plane bucklings were used throughou preferencn i t e

to core-height-average bucklings. This implies that the fluxes quoted are for the

mid-height plane, and they have been so normalized.

The reactivity scale used requires some explanation. Instead of the normal

definition of % reactivity i.e. p = 100 (1 - l/k _,), it is practice with HIFAR

operationa a " %e l reactivitydatus o at equivalene 1 whicth p" , s i h t reactivitn yi

a core containing 3.2 kg of 23SU. This has the advantage that the change in p' due

to a rig, making a fuel change, or moving the control arms is independent of the

fuel loading relationshie Th . p use o derivt d e calculate valuedp s i s 1

p1 - p (M/3.2)0'7

where M is the reactor 235U loading in kg. In comparing fuels of various enrichment,

however s necessari t i , o alloyt r thewfo 239Pua e 235s th welU s a loadingl o T .

account for the higher cross section of 239Pu, twice the 239Pu mass has been added

to the 235U mass to obtain an 'equivalent 23SU' mass. The reactivities given in

the preceding paragraphs are p1 values derived this way.

The reactor models outlined above have been checked by comparison with the

end-of-cycle core particulaa stat r efo r HIFAR operating program (No. 251)n I .

making this comparison, the core state calculated using the simple fuel accounting

program HIBURN (McCulloc Trimbld an h e 1969 s usede exces)wa Th . s reactivitp y 1 e controth hel n di l armcalculates wa s 1.49e b o %dt usin e explicigth t burn-uf po

individual fuel elements and 0.84% using a uniform core burn-up, compared with 1.43%

actually observed.

The basic results calculated here are for cores of uniform burn-up. This

approach has been used for simplicity, and its accuracy is considered in Section 4.

3.0 Basic Results

HIFAR fuel cycle data averaged over the last 21 programs are presented in

125 Table 2. It is evident that the rig burden is currently low, and the fuel burn-up correspondingly high. The fuel management scheme involves loading most new elements

neacore rth e centr movind ean g elements outward (afte least delara a e cyclf tyo on e ou reactof to cooling)r rfo .

burdeBecausg ri e f HIFAno e varth y Rma y considerabl e futureth n yi have ,w e

chosen to consider two cases which cover the likely range. The first is based on the prograf o averagcord n whic1 a en me 25 d e hstatha eth t burn-uea 45.1f po 6 MWD/

element and an in-core rig p' of 2.3%. The second, also for standard HEU fuel, has a burn-up of 33.68 MWD/element and an extra in-core rig p1 of 3.9% to give the same

k . The extra rig burden was included via a uniform thermal absorber in the 20 elements not already having an explicit rig loading.

The results obtained from XY calculations using bucklings from the appropriate

RZ calculation are given in Table 3. The p' values derived from these cases are

summarized in Table 4. The columns labelled high and low burn-up give the excess p'

values for the two situations. They show : chosee th (al n) al reduce d enrichment cases hav similaea higher ro r

reactivity than standard fuel over this burn-up range;

reactivite th ) (b y performanc enrichmenw lo f eo t fue bettes li higt ra h burn-up than at low burn-up. Additional calculations for the low burn-up cases were made without the extra

rigs included resulte th d ,san use giv o dratt e eth changf e o wit p f heo burn-up 1 presented in Table 4. These figures can be interpreted as the rate of reactivity

loss (excluding transient effects) within a program, and demonstrate that the control

require balanco dt e thi enrichmensw lo los smallee s si th r trfo fuels. The axial form factors for the thermal flux and the average core centre-plane

thermal fluxe casesl givee thermaal e sar Tablr Th n .ni fo e5 l flux used throughout

Westcote isth t flux reductione Th . corn si e % therma13 d an l % fluaboue 20 xar , t7%

C fue20 fo lrd respectivelyan 45A A ,20 . Further informatio flue th x n nchangeo n si

the high burn-up case gives si Figuren wher4 ni o t e1 s. fluxe s alon X-axie gth r sfo 80A fuel and flux ratios for the other fuels are given. In these plots the fast . epi-thermae eV th 1 d flu 1. an >0.s xi o V t 8 Me l V fluke fros x1 i m9.

126 A selection of reactivity coefficients and other safety related parameters

which have been calculate perturbatioy db n theor comparee yar . 6 Table n d i d an s5 The fuel coefficient for the central element gives the change in Sk/k per g of

U burnt and includes all long term burn-up effects. The absorber coefficient 235 varies considerabl thit ybu s mainly results frochange mth fuen ei l loadin therd an g e

is a much smaller variation in values of p" per cm . The prompt neutron lifetime,

i, also shows some variation with fuel loading e smalTh . l changeeffective th n si e

239 delaye variatiode th neutroe o Pth t u n e ni ncontribution du fractione ar , ff ,ß . Only one set of values is given for the remaining coefficients in Table 6 as the

variation with burn-u less pi s tha. Therconsiderabla n3% s ei e increase th n ei

fuel temperatur wit) e °C coefficienh r decreasinpe k (Ôk/ t g enrichmen thit tbu s

coefficient is of minor importance. The fuel plus coolant temperature coefficient

is the coefficient for constant coolant density and must be combined with a density

change and the void coefficient to obtain a total temperature coefficient. The

temperature coefficient and void coefficient (

C r fuelvaluee 20 fo Th .80d % Aan s obtaine10.5 d an changer werdfo % ' p e9.2 n s i

80A and 20C respectively and the worths of the first 0.2 m of top reflector were

6.6 and 7.6% respectively.

The power distribution within the fuel element for the various fuels is given

izerresultnr e Tablfo Th o e distributionburn-u. e es7 ar th d pan s flatten

slightly as burn-up proceeds. The results are practically identical for all cases.

Plutonium production figure givee ranga sar Tablr n burn-uf ni eo fo e8 p which

covers the anticipated discharge values. 0 Reactor-Follo4. w Calculations To supplemen informatioe tth n obtained from compariso coref no uniforf so m

burn-up, some limited reactor-follow calculations have been done. These have not been performed by the direct use of the general purpose reactor physics methods

used in the preceding sections,but with a program similar to the HIBURN program

whic uses HIFAr hi fo d R fuel management. This program, HIFUEL, whic currentls i h y still being developed, retains the HIBURN approach of using flux factors and

127 reactivity coefficients, but uses calculated values throughout rather than experimentally based data included ,an treatmensa saturatinf to g fission products.

The comparison has been made over the last 21 programs which gave the average fuel cycle data of Table 2. The actual fuel movements and time at power have been followe witt dbu constanha shuy da t t3 dow n between programs program7 e Th . s

before this were also followed but have not been included in the averages presented.

The standard fuebees lha n compared only wit fuel,whicA h20 mose th t s hextremi e

case.

A summary of the results is given in Table 9. Case 1 represents the actual HIFAR situation. Case 2 represents a situation with a high rig burden in which the extra 4% in reactivity has been obtained by increasing the fuelling rate by

50%. A 50% increase was chosen as this could be modelled by maintaining the same

average cycle lengt makint hbu evern gi cyclesy2 tthe fuel changes tha casn i t 1 e powee twicd caswern bee s th I an cas e r3 ha e eenW th . e1 M mad3 raise 5 n 1 ei o dt

refuelling rate used (again chose simplicity)r fo n , which resulte 2.1n di % extra

reactivity using standard fuel. Case 3 has been included because HIFAR uprating maconsideree yb futura s da e option.

The results show that : cyclf o frod chang' e ep en th m n standarei ) n i i A fue 20 s l i o dt

each case about 0.2% above that inferred frocorrespondine mth g

uniform core calculation, and

ii) the use of 20A fuel has little effect on the total change in p1

oveoperatine rth g program. Thi becauss si smalleea r lose sdu o fuet l burn-u compensates pi largea y db rsaturatino t los e sdu g

fission products. It is inferred from these results that the uniform burn-up results may be used

with confidence for all fuel types, and that operation at 15 MW introduces no

untoward additional difference between fuel types.

128 5.0 Conclusions

The performance of HIFAR 80% enriched 150 g 235U fuel elements is reasonably well matched in terms of reactivity by :

enriche% a)45 d 235U element ~1.t a sl 0containin-A L UA 23Sg s U5 a g15

chango N gern" . fuen U ei 3 l element geometr requireds yi e fueth l d ,an

technolog differenn (i withis y i e range us n th t n ei geometries somn )i e other high performance reactors.

b) 20% enriched 23SU elements containing 170 g 235U as UAL -Al at ~1.6 X gem"3 U. This density is at the upper limit of currently fully proven

technology, and would require a change in fuel element geometry (thicker

plates) which would reduce thermohydraulic safety margins.

23S 235 c) 20% enriched U elements containing 160 g U as U3O8-Al at ~2.3 gcm~3 U. No fuel element geometry change is entailed, but the fuel technology

still remains to be fully developed and proven.

The largest performance penalty in using reduced enrichment fuel comes from the

reduced core thermal flux, and is substantial at the 20% 235U level. There are no

particularly significant changes in safety related neutronics parameters.

0 Referenc6. e McCulloch, D.B. and Trimble G.D. (1969) - A method of estimating fuel burn-up

and higher isotope production in the reactor HIFAR. AAEC/TM508.

129 TABL : EFUE1 L DESCRIPTION

Fuel type Fuel meat Enrichment 235U/element Uranium density Fuel meat wt% (g) (gcm~3) thickness (mm)

80A U-A1 Alloy 80 150 .535 .66

45A UAl -Al 45 155 .984 .66 X 20A UAl l-A 20 170 1.60 1.00 X

2 OB U308-A1 20 170 2.43 .66

2 OC U308-A1 20 160 2.28 .66

TABL : EAVERAG2 E FUEL CYCLE DATA

In-core rigs ' ,p 2.79%

Reflector rigs 1 ,p 2.19% Excess p1 at shut down 1.64%

W M 0 1 Power Energy per program 227.8 MWD

New fuel elements per program 3.52

Discharge burn-up per element 64.7 MWD Average burn-u elemenr ppe shut ta t down 40.D 5MW

235U loading at shut down 2.48 kg

130 TABL ; CASE3 E DESCRIPTIO kefD NAN f

Fuel type Fissile Burn-up In core K Reactivity ef «f mass MWD/element rig burden (P%) (kg) (P'%)

80A 2.3611 45.16 2.32 1.01054 1.04

80A 2.7144 33.68 6.22 1.01047 1.04

45A 2.5953 45.16 2.32 1.01626 1.60

45A 2.8937 33.68 6.22 1.01275 1.26

20A 3.0470 45.16 2.32 1.01820 1.79

20A 3.3583 33.68 6.22 1.01047 1.04

2 OB 3.0473 33.68 6.22 1.02708 2.64

20C 2.7939 45.16 2.32 1.01564 1.54

20C 3.1061 33.68 6.22 1.00880 0.87

TABL : REACTIVITIEE4 S

Fuel type P'% A (p '%) /A (MWD/element) High burn-up Low burn-up

80A 0.84 0.92 0.347

45A 1.38 1.17 0.322

20A 1.73 1.07 0.287

20B 2.73

20C 1.40 0.85 0.294

131 TABLE 5

Centre Element Thermal Flux Fuel coeff. Absorber Centre- Axial I 6eff Case —6k per g coeff. plane form _6k average factor (sec) of 23SU burnt cm2

High Burnup 80A -2.066x10"* -2.946x10"" 1.113X101" 1.044 6.44xlO~" 6. 92x10" 3

45A -1.8 31x10" "* -2.763x10"" 1.033X101" 1.038 6.24x10"" 6. 82x10"3

20A -1.438x10"** -2.370x10"" O.SSOxlO1" 1.022 5.80x10"" 6. 70x10" 3

20C -1.596x10""* -2.554xlO~" 0.952X101" 1.031 5.98xlO~" 6.68xlO~3 I

: Low Burnup 80A -1.814x10""* -2.635x10"" 0.976X101" 1.037 5.90x10"" 6. 94x10" 3

45A -1.647x10"'* -2.500x10"" 0.920X101" 1.032 5.76x10"" 6. 86x10"3

20A -1.330x10""* -2.202x10"" 0.803X101" 1.017 5.44x10"" 6. 76x10" 3

20C -1.452x10"** -2.352x10"" 0.862X101" 1.026 5.58x10"" 6. 7 5x10" 3

TABL ; REACTIVITE6 Y COEFFICIENTS (ök/fuer pe kl element)

Core Average Centre Element Fuel Fuel Fuel & Coolant Fuel Fuel s Coolant Type Temperature Coolant Void Temperature Coolant Void Temperature Temperature

80A -0.4xlO~7 •3. 6x10" 6 -3. 5x10" s -0.6xlO~7 -5. 4x10" 6 -5. 4x10" 5

45A -3. 1x10" 7 -3.7xlO~6 -3. 4x10"5 5.4xlO"- 7 -5. 4x10" 6 2xlO. -5 ~ 5

20A -6.9xlO~7 -3. 4x10"6 -3. 0x10" 5 -12. 5x10"7 9x10. "-4 6 -4. 6x10" s

20C -7. 2x10" 7 -3.8xlO~6 -3. 2x10" 5 -12. 8x10" 7 -5.6xlO~6 -4. 9x10" 5

132 TABLE 7: POWER FRACTION Bï FUEL RING

Power Fraction Fuel Type Annu1 s lu Annulus 2 Annulus3 Annulus 4

80A 0.194 0.225 0.266 0.315

45A 0.198 0.227 0.264 0.312

20A 0.197 0.226 0.263 0.313

20C 0.197 0.226 0.264 0.313

TABL : PLUTONIUE8 M PRODUCTION Plutonium Burnup Isotopic composition (wt%) production Fuel Type MWD/el 239 2 g/el Pu *°Pu 2*lpu 2*2PU

80A 45 0.58 82.8 13.2 3.6 0.3

55 0.66 78.5 15.8 5.2 0.6

65 0.73 73.9 18.2 6.8 1.1

45A 45 1.93 83.6 12.7 3.4 0.3

55 2.22 79.5 15.2 4.8 0.5

65 2.46 75.3 17.5 6.3 1.0

20A 45 4.20 84.9 11.7 3.1 0.2

55 4.89 81.5 13.8 4.3 0.4

65 5.51 78.0 15.8 5.5 0.7

20C 45 4.09 84.2 12.2 3.2 0.3

55 4.75 80.5 14.5 4.5 0.5

65 5.32 76.8 16.6 5.8 0.8

133 TABLE 9: REACTOR FOLLOW - AVERAGE RESULTS

Case 1 Case 2 Case 3

Power (MW) 10 10 15.

Refuelling rate (elements/program) 3.52 5.28 7.04

Results using standard fuel:

p1 increase c.f.) Cas(% e1 - 4.00 2.14

p) 1 (% los cyclr spe e 5.71 6.02 7.50 Results comparin witA 20 gh standard fuel:

change in p' loss per cycle {%) +0.08 +0.07 -0.09

) (% f cyclo changd ' ep en n ei +0.80 +0.07 +0.30 (+0.59)f <-0.18)f (+0.15)f

t Correspondin cyclf o derive' d ep gen d from Tabl. e4

134 Flg . 1 Fluxe. t Cora s e hld-pl-ane

l.O • i • 1 ' 1 ' 1 • i i '_ Core Graphite ' »^ D°°

1.2 \ - t ---J^^' • o "'N \ N \ ^ \ . 0.8 \ - \ \ —— na \ —— cri-nom ~ \ —— nom." \ v x n L. \ s . \v . LÜ N - *"—• • -x^ % ^^^*- ~ n n i 1 i S-»-^ T"-- ^i_ L ""~~~~— • 0.0 t»0.0 80.0 120.0 X. - QX\ s Com3

Fig . 2 Flu. x Ratio f Coro s e Nld-plane £t5fl / 80fl Fuel

Core D20 Graphl te _

- 1.00 4- ccO

f 0.96 —a n — —— cpi-nenc —— nom.

0.92 *-= 0.0 1*0.0 80.0120.0 - Xaxl s Com3

135 Fig. 3. Flux Ratios at Core Mid-plane 20fl / 80fl Fuel I.IO • 1 ' i • i • i \ • Core D,0 Graphite 1.05 _ o ^ —r..-»- — +. ^~— ^ ^ 0.95'^^ X +~'~'~ • f LU s ' —— run 0.85 - / —— eri-wam ^ ———— TKJWC _ " ——--"S : ft 7C "~~7~ i . i.i.i. i . 0.0 i+0.0 80.0120.0 - Xaxl s Com3

. Figi» Flu. x Ratio t Cora s e Mid-plane 1.10 20 C/ 80f l Fuel

Core D20 GrapJ e t h I

1.00 ooc

0.90 na

0.80 0.0 £+0.0 80.0120.0 s Ccnl ax O - X

136 APPENDI PAPE- XC 2 RC

AN ANALYSIS OF COOLANT FLOW STABILITY IN DIDO TYPE REACTORS

T.M, ROMBERG Nuclear Technology Division Australian Atomic Energy Commission

1. INTRODUCTION

The efficient removal of heat from the core of a DIDO type nuclear reactor depend steade th n yso flo coolanf wo t throug componene hth t fuel

channels. Under certain operating conditions coolane ,th t flow througe hon or more fuel channels may become unstable; that is, depart from the desired

flow rat anotheo et r stable (but less favourable) flow rate condition. This phenomeno knows floa ni s wna excursio excursivr no e flow instabilityd ,an can lead to dryout or even burnout problems. This paper discusses how the stability limits of DIDO type fuel channels can be assessed, and presents multi-annularesulte th r sfo r fuel channels use HIFAn di R (High Flux Australian Reactor).

2. MECHANISM OF EXCURSIVE FLOW INSTABILITIES

The mechanism of excursive flow instabilities is best explained by reference to the steady state pressure drop - flow rate characteristic of a typical fuel element shows ,a wits Figurn A ni h . othee1 r reactor types, DIDO type reactors have core configurations in which several fuel channels are connected hydraulically in parallel between common plena located at the inle outletd tan . With this configuration externae ,th l pressure differential betwee plene nth a (i.eavailable th . e pump hea supplr do y curve approxs )i - imately constant independens i ? t i coolane tha , th tis f to t flow conditions

137 in any one channel. However, the pressure drop (or demand curve) generated floe bcoolanf yth wo t throug passagee hth fuea f lso channel depende th n so

operating conditions (power input, coolant inlet temperature and pressure, flow rate, etc. geometricad )an l dimensions.

As Figure 1 shows, the sequence of events which leads to an excursive

flow instabilit fuea n lyi channe mor, lor e importantly coolana n ,i t passage

within a fuel channel, may be described as follows :

(1) For a given power input, inlet temperature and pressure, a

typical demand curve consists of an all-liquid region of

positive uppe e slopth rn ei flo w rate range ,two-phasa e

(steam/water) regio negativf no e slopeall-stean a d ,an m

region of positive slope in the low flow-rate range. For

normal operation, the minimum (Ap ) in the pressure drop-

flow characteristic coincides wit onsee hth subcoolef to d

boiling, and thus the correlations used in computer codes

to predic onsee tth subcoolef to d boiling hav importann ea t

influence on the results.

(2) For a particular coolant passage, the operating point is

defineintersectioe th y d b deman e th f ndo characteristic

Ap(w) with the supply characteristic (Ap ), normally on the

all-liquid characteristic operatinn A . g negativ e pointh n ti e

slope region (e.g powet .a r unstables inpui ) tPS d flo,an w

excursions will proceed to intersection points on either

the all-stea all-liquir mo d characteristics.

(3) Flow excursion powee th rd slevean occu, t la p A r > whe p nA m s this condition is denoted as the instability threshold power

(ITP) ITPe Th ., normalised with respec average th o tt e

operating power denotes instabilit,e i th s da y power ratio

(IPR)-

138 3. STABILITY MAPS

Two computer codes (FLEX, LOCO) have been developed at the Australian

Atomic Energy Commission (AAEC asseso )t stabilite sth y HIFAlimite th Rf so

multi-annular fuel channels. The purpose of these codes is to predict the

demand characteristic individuaf so l annular passages withi fuena l element

pressure locue anth dth f so e drop minima with variation powen si r inpud tan coolant inlet temperature. If the pressure drop minima shown in Figure 1

are plotted as a function of power input for various coolant inlet temperatures, then Figure 2 is obtained. The instability threshold powers (ITPs)

varioue atth s coolant inlet temperatures correspon pointe th o sdt where eth

suppl pressure-drod yan p minima characteristics intersec. ) p A t= (i.e p A .

Plotting the ITPs as a function of coolant inlet temperature gives a stability

map of the fuel channel, as shown in Figure 3. These stability maps provide the requisite information for comparing the stability limits of various fuel

channel designs.

4. HIFAR FUEL ELEMENT FLOW STABILITY

principae Th l dimension HIFAe th Rr Marsfo multi-annula5 k4/ r fuel

element are given in Figure 4. The stability limits for a range of channel/

fuel/cladding dimensions, includin standare gth d fuel element, were calculated

using the LOCO code, and are presented in Figure 5. The results are based followine onth g assumption: s (1) the pressure differential between the inlet and outlet of

the annular coolant gap a 26.(3.8s si kP 54 psi);

coolane th ) t(2 inlet temperatur; °C 0 8 o t e 0 rang5 s ei

secone th coolan (3e f d)th o outep tga r annululease th ts si

stable annulu e fueth l f so elemente th uses i s da d ,an

referenc assessinr fo e fuee gth l element stability limits;

(4) the fuel element/annulus power ratio is 3.554, and does

not vary between cases.

139 The instability threshold power (IIP powed )an r ratio (IPRe b y )ma plotted as a function.of coolant gap width at .a coolant inlet temperature of 50°C shows ,a Figurn ni . eThi6 s Figure shows that their relationship is linear e equationgives i th y d nb , an s

, 5 53 - IT g P 6 (kW85 )= or IPR = 0.9011 g - 0.5629 , where ITP = 950 IPR and the coolant gap (g) is in mm.

5. REFERENCE N 'FLEXSO D 'LOCOAN * ' CODES

(1) Romberg, T.M. Multi-annular Fuel Element Boiling Stability Eng. M . .

Sei. Thesis, Universit N.S.Wf yo . (1972).

(2) Romberg, T.M. Noise Analysis of Coolant Dynamics in Boiling Two-phase

Flow Systems, Ph.D. Thesis, university of N.S.W. (1978).

(3) Romberg, T.M., Rees, N.W. Multivariate Hydrodynamic Analysis of Boiling

Channel Flow Stability from Inherent Noise Measurements. Int.

Journal Multiphase Flow, 6(6), pp. 523-551. (1980).

{4) Romberg, T.M. LOCLineariseA O- d Mode Analysinr lfo Onsee g th Coolan f to t

Oscillations' and Frequency Response of Boiling Channels. AAEC/E

Report (in preparation).

140 /•»-All steam characteristic Increasing Power

l liquiAl d (Zero Power) characteristic Commos n (supply) pressure - — differential Ap

Pressure drop minimum Apm

l^-Operating point Flow Rat) e(W FIGURE 1. FLOW-PRESSURE DROP CHARACTERISTICS (constant inlet temperature, pressure)

E a. Coolant Inlet Temperatures 8O°C 7O°C 6O°C 5O°C

ç 2 Unstable I Stable

a. Instability Threshold Powers (ITP) Power Input FIGURE 2. POWER INPUT-PRESSURE DROP MINIMA CHARACTERISTICS'

MAXIMUM RATED FUEL ELEMENT

Unstable Region o QC 1 O W a R P I -IT r o P I % U) Stable Region ÜJ

Coolant Inlet Temperature (°C) FIGURE 3 COOLANT INLET TEMPERATURE—POWER MAP

141 STANDARD 15O 5 qU Oute r Tube Tube Tube Tube Tube 4 3 21 8O°/o U.V. 17Wt°/n i oU U-Al ALLOY Ê CD S

.3 •C7

t_ 6J

W Cd 2.923 2.635 2.635 2.635 2.9O3 CASE 3 4——————— ^ 4 ——————— »» g V c

J^ 62.31 I 72.11 81.92 91.72 * 1O1.4 4 —————

FIGURE 4. HIFAR MARK 4/5 FUEL ELEMENT DIMENSIONS

142 -24OO Pressure Drop = 26.5 k Pa -23OO or S o Flex Code - Standard Case LU -22OO ex. o Loco Code LU -21OO_ O CL - 2000^ Lu2.O -19OO ce i LU Lu Lu _l -18CO n Lu LU LL -17OO H LU LU -1600g LLl s -15OO LU 1.5 8 -14OO LU -130O ,_ z -12OO y -11OO | STABLE S -1000 S 5c 1.0 o -90O Q. -800

O 8 O 7 O 6 5O COOLANT INLET TEMPERATURE (°C) FIGURE 5 EFFECT OF COOLANT INLET TEMPERATURE ON" THE INSTABILITY THRESHOLD POWER FOR VARIOUS GAP WIDTHS

143 HIFAR EFFECF TO COOLANT GAP WIDTH ON INSTABILITY POWER

Coolant Inlet Temp.=-5O°C : Pressur e Dro 26.a p= 5kP

a Flex Code o Loco Code

Standard UNSTABLE ^ Case 1

= 1 2.6 2.7 2.8 2.9 3.O 3.1 32 33 COOLAN WIDTP TGA H (mm)

FIGUR . EEFFEC 6 INSTABILIT E COOLANTH F TO N O P Y TGA THRESHOLD POWER

144 APPENDIX D Enrichment Reduction Calculations for JRR-2 Reactor, JAERI (Japan)

Y.Naito, F.Sakurai, M.Kurosawa, Y.Torii

Japan Atomic Energy Research Institute

145 I. Nuclear Calculatio JRR-f no CorU CoreU LE 2Cor HE eU d ,ME e an Y.Naito, M.Kurosawa, Y.Torii 1 Introduction JRR-2 is a 10 MW research reactor moderated and cooled by heavy wate fueU rE lwit% element3 h9 core .Th e typR consistMT e 0 2 fue f so l and 4 cylindrical type fuel elements. The current core configration are show Fig.n ni Fig.2d an l . Fuel element showe s ar Fig.3n ni . Before startin reducee gth d enrichmen d talread ha program d yha e ,w the program of using only cylindrical fuel elements in the progressive corU extenHE o et fuee dth l cycle mako lengtt ed han compensatio r fo n the reactivity worth of the increasing irradiation mateials. Because, the cylindrical fuel element has the lager reactivity worth than the MTR type fuel element at the same U-235 loading per element, and the former has the tube inside it for irradiating materials. Therefore onle intene ,w us yo t dcylindrica l type fuel elementn i s the reduced enrichment cor f JRR-o e o providt 2 lagee eth r excess reactivity (about 2 %AK/K) and to extend the fuel cycle length from about 600 MWD to about 800 MWD. Our enrichment reduction program of JRR-2 does not include any chages of the geometry of fuel element nor the chemical form of fuel meat so as to make easy to pass the licensing procedur reductior efo enrichmenf n o years w fe n a thes.I n ti e condi- tons, the U-235 loadings per fresh fuel element of MEU and LEU core were estimated. In section 2, calculation method and models and in section 3, computed e measureth result d an ds dat showne aar section .I e th , n4- conclusion of the enrichment reduction program of JRR-2 is described.

2 Calculation method and models 2.1 Outline of RETER-ACE for analysing research reactors at JAERI A code syste s beeha mn developed with KURRI (Kyoto University Reserch Reactor Institute r analysinfo ) e corth ge performancf eo reserch reactors. This code system consists of three parts. The first o obtait e parmulti-grous th i nt p nuclear constants library (MGCL) which is generated from the nuclear data file ENDF/B-4(Ref.1), and the second part is to obtain burn-up dependent cell averaged few group constants table (FG-table) by using the SN code ANISN-JR(Ref.2). The thir dcalculato t par s i t burn-ue eth p dependent core performance using the three- or two-dimensional neutron diffusion code FEDM(Ref.3) or 2DFEM. 2 Calculatio2. n models Models used for FG-table preparation were chosen so that cell averaged cross sections are indentical with the cell averaged cross sections obtained by detailed model calculations(in the case of cold clean core). The models used for FG-table preparation are shown in Fig.4 detailee ,th d model showe sar Fig.5n ni . A 1/6 core model used for FEDM-BURN calculation is shown in Fig.6.

146 JRR-2 current cor cylindrica4 e typR d consistMT ean 0 2 f so l type fuel elements(U-235 loadin 195f o gelement)r gpe thin .I s calculation(Case 6 cylindrica cord e 1)an th , e R consistMT 8 l1 f sfueo l elemente ar s chose expreso t n core sth e witcor6 h1/ e model difference .Th Keff eo f between the former and the latter is small. It is confirmed by the experimen t JRR-2a t . Excep core Casr th etfo , e model1 s were computed consist of 24 cylindrical type fuel elements. Computed cases are shown in Table 3. 3 Computed Results Calculations were done for caseo tw , s wit fuel(U-23U hHE 5 loadinf go 195g per element, the current core and the progressive core consists of only cylindrical fuel element) caseo ,tw s witfuel(U-23U hME 5 loading of 192g, 220 element)r gpe case ,on e witU fuel(U-23LE h 5 loadinf o g 237g per element). See Table 3. 3.1 Compariso effectivf no e multiplication factors computey db the FEDM-BURN with measuree don The burnup dependent effective multiplication factors calculated by the computer code are shown in Table 2 and Fig.7 with measured ones. e measureTh d excess reactivit s obtainewa y d witcontroe th h l rods critica lcontroe heighth d ltan rods worth curve measure y Positivb d e Period Method. The excess reactivity obtained is shown in Table 1. The computed results are shown that calculated values are about 2-^3 % smaller than measured ones at low burn-up. Except for the range, the effective multiplication factor obtaine y FEDM-BURb d N show goosa d agreement with measured one. thinke W measuree ,th largee wortd th dro s ri h e thareath nl one, the higher excess reactivity because of the interaction among rods. It e calculate reasoth e th y seeme nwh b o dst value e smallear s r than measure dburnuw onelo t spa show Fig.7n ni . 3.2 Calculations with HEU, MEU and LEU fuel cores To check the propriety of reactivity worth of reduced enrichment fuel of JRR-2, effective multiplication factors were calculated. The computed result e showexcesL Fig.10d ar n sEO Tablan i ne s 3 .Th e reactivity of Case 2(HEU progressive core) is about 2 %AK/K higher than Case 1(the curren texcesL EO core) e s Th reactivit. Casf yo e 3(U-235 loadin f 192g/MEgo U fuel element e almosar ) t e th sam s a thaef o t current core. These calculation fuee th ldon e r cyclsar e fo e lengtf ho MWD0 60 . excesThL eEO s reactivitie Casf so e 4(U-235 loadin 220g/MEf o g U fuel element Casd )an e 5(U-235 loadin 237g/LEf go U fuel element %AK/2 e )Kar lager than the current core. These calculations are done for the fuel cycle length of 800 MWD. 3 3. Neutron flu burnud xan p characteristifuelU LE sd an U ME f co Thermal(<0.683 eV) and fast(>183.2 KeV) flux distributions(0-X showe ar n Fig.ni L d Fig.9EO sectioan 8d an .Fig.6 n L ni FasBO t )ta fluxes of the MEU core(Case 4) and LEU core(Case 5) are almost same value as that of the HEU current core(Case 1). Thermal fluxes of flux smalle% e 0 MEU 2 th corU U abou, e trarLE % t ,HE ear a 5 tha p1 e t th n current core.

147 The EOL contents of U-235 and Pu(total) in each fuel are shown in Fig.11, Fig.12 and Fig.13. 4 Conclusion As describing in Section 1, it is required for the reduced enrichment cor reactivitL eEO thae th t s aboui y 2 %AK/t K lager e thath n current core, and that the fuel cycle length is extended from 600 MWD to 800 MWD. The computed results in Table 3 and Fig.10 indicated that the U-235 loadings of 220g/element with MEU fuel(uranium density of l.SQg/cm3 ), 237g/element with LEU fuel(uranium density of 3.85 g/cm3 ) are sufficient for the above conditions. The uranium density of 1.59g/cm3 is achieved for UAlx-Al fuel,the

uranium density of S.SSg/cm is achievable for only U3Si-Al fuel. According to thermal-hydrauli3 c analysis of JRR-2 by F. Sakurai, the the conversion into LEU fuel with the meat thickness of 1.0 mm is feasible thed .An n recentl hige yth h uranium density s fuebeeha ln developed. Therefore, it may be feasible to convert into LEU fuel with change geometre th f so fuef yo l element and/o chemicae th r e l th for f mo fuel meat. However, we intend to achieve the enrichment reduction of JRR-2 without any changes of the fuel design except for uranium density, in order to make easy to pass the licensing procedure for reduction of enrichment. Conclusively, the core chosen as the reduced enrichment core of JRR-cylindrica4 2 s thai 2f o t l elements with MEU(4 5Enr.% ) UAlx-Al fuel U-23,a 5 loadin 220g/elementf go .

148 II Thermal-Hydraulic analyses of the JRR-2 F.Sakurai The present core consists of 20 MTR-type fuel and 4 cylindrical type fuel elements showFigurcore . Th in ne . econfiguratio3 shows ni n ni Figure 1. JRR-2e casth e I f nth specificatio,a eo fueU s estiLE lwa e th - f no mated roughl y neutronib y d thermal-hydraulican c analyses(Ref.4)n .I these analyses, it was assumed that the LEU fuel core consisted of 24 cylindrical type fuel elements, the uranium density of this LEU fuel is meae 4 g/cmth 2. t presene . d (Basethicknes3th ,mm an n 0 do t 1. fues si l technology, it seems to be difficult to fabricate curved fuel plates with the meat thickness of 1.0 mm). Thus, the thermal-hydrau- lie analyses with COBRA-X/RERTR were performe core th e n witdo h thesU eLE cylindrical type fuel elements. 1 Thermal-Hydraulic Calculations Correlatio datd nan a usen thermal-hydraulii d c calculatione ar s summarized in Tables 4 and 5, respectively. The coolant velocity for corU th increaseeLE s ei 4.0o t d 3 m/sec froU mHE 3.8e th 0 m/sen i s ca core. Figur show4 e1 heae sth t flux distributio JRR-2e th n .ni 2 Results summariz6 Tabl1 Figure , d seee e6 b an resultse n ni 5 n eth s 1 ca s .A maximue th , m16 Figure fued an l 5 surfacs1 eU LE temperatur e th r fo e case is lower than that for the HEU fuel case. Thus, the LEU fuel core coreU HE thaB .larges e DN nha th d ran margiB ON o nt 3 Conclusions The margin to ONB and DNB for the LEU fuel case are larger than that U fueHE l e increaseo casefot th re sdu d coolant velocity. Therefore, based on the results from the thermal-hydraulic analyses the conversion intU fueoLE l witmeae th ht appearthicknesm m feasiblee 0 b 1. o st f so .

References 1. ENDF/B Summary Document BNL-NCS-1754 edid 12n . (1975) 2. K.KOYAMA, et., "ANISN-JR, A One-Dimensional Discrete Ordinates, Code for Neutron and Gamma-Ray Transport Calculations", JAERI-M 6954, (1977) 3. Y.NAITO, et al., "A New Mixed Method with Finite Difference and Frinite Element Method for Neutron Diffusion Calculation", J. of Nucl. Sei. & Tech. Vol.18, No.8 (1981) 4. "Report on Phase A, ANL-JAERI Joint Study on the Use of Reduced Enrichment Fuel in the JAERI Research Reactors" JAERI,August,1980.

149 Table 1 Measured excess reactivity vs burn-up of JAR-2

Burn-up (MWD) Reactivity (ZAk/k)

0 15.03

32.7 9.99 90.0 9.09 140.2 8.40 200.0 7.58 249.6 6.96

300.0 6.72 360.6 6.34 420.0 5.66

471.6 5.11 530.0 4.85 582.3 4.31

Moderator Temperature 20°C Reactor thermal power 10 MW 1976-y DatMa ^e: Oct. 1976

150 Tabl 2 e Burn-up dependent effective multiplication factors of JRR-2

Burn-up Measured Calculated (MWD) ZAk/k K«ff Keff

0 15.03 1.177 1.1510 32.7 9.99 1.111 50.0 1.0943

90.0 9.09 1.100

140.2 8.40 1.092

200.0 7.58 1.082 1.0809

249.6 6.96 1.075 300.0 6.72 1.072

360.6 6.34 1.068

400.0 1.0621

420.0 5.66 1.060

471.5 5.11 1.054

530.0 4.85 1.051 i 582.3 4.31 1.045 600.0 1.0417

151 tL/ol Table 3 JRR-2 U-235 Loading with Uranium Enrichment of 45% and 20% Case 1 Cas2 e Case3 Case 4 Case5 Case6 Enrichment, % 93 93 45 45 20 20 Fuel Type U-A1 Alloy U-A1 Alloy UA1 -Al l -A 1 UA l -A 1 UA UA1 -Al X X X X Fuel Geometry 18 MTR Type 24 Cylindrical 24 Cylindrical 24 Cylindrical 24 Cylindrical 24 Cylindrical 6 Cylindrical

Burnup.MWD 600.0 600.0 600.0 800.0 800.0 600.0 Meat Thickness 0.51 0.51 0.51 0.51 0.51 0.51 ,mm

Clad Thickness 0.38 0.38 0.38 0.38 0.38 0.38 ,mm

U-235/Elemenc t 195 195 192 220 237 206 -g U-235 Density 0.62 0.63 0.62 0.72 0.77 0.67 ,g/cm3

Uranium Density 0.66 0.68 1.39 1.59 3.85 3.41 ,g/cm3

BOL Keff 1.1510 1.1715 1.1509 1.1790 1.1701 1.1452

EOL Keff 1.0417 1.0608 1.0414 1.0578 1.0575 1.0429

; JRR-2 current core consists of 20 MTR Type and 4 Cylindrical Type Fuel Element. *;Based on a power level of 10 MW. ';U-23 e fresth hr 5 fo fueetc e l ar .element . Table 4 Correlations Used in Thermal-Hydraulic Calculations for the JRR-2

Correlation

Friction Factor f - 0.316 Re~°*25

(Re)JRR-2 "•30 '00° Single Phase Heat Dittus-Boelter Transfer Coefficient Nu - 0.023 This correlation cannot take into account the temperature filme th ris .n ei

Subcool Boiling Heat Bergles-Rohsenow Transfer Coefficient

Critical Heat Flux Labuntsov

Tabl 6 e Thermal-Hydraulic AnalysiJRR-e th 2r fo s

HEU Fuel LEU Fuel

Fuel Plate Thickness (cm) 0.127 0.176 Water Channel Thickness (cm) 0.30 0.25 Water Channel Area/Element (cm ) 40.82 38.23 Reactor Thermal Power (MW) 10 10 Primary Flow Rate (m /min) 22 22 Core Inlet Water Temperature (°C) 57.1 57.1 Core Outlet Water Temperature (*C) 67.5 68.9 Clad Surface Temperature (Maximum) (°C) 115.8 111.9 Water Velocity (m/sec) 3.8 4.03 Average Heat Flux (W/cm) 31.7 31.9 Margin to ONB 1.34 1.46 MargiB DN o t n 11.6 11.9

153 Table 5 Data Used ln Calculations for the JRK-2

HEU Fuel. LEU Fuel

Humber of Plate« IS 15

Heat Thlckne*« (on) 0.51 1.00 flat* Thlckne«» (ao) 1.27 1.76

Active Fuel Length (mm) 600.0 600.0

Active Fuel Width (mo) A8.6 K 3, 57.6 x 3, 66.5 x 3 48.6 x 3, 57.6 x 3,-66.5 x 3 3 75. x 484. 3 4x 3 75. x 484. 3 4x

Coolant Channel Thlckne«« (na) 3.0 2.5

Coolant Channel Aret (DO?) 151.3 x 3, 178.1 x 3, 204.9 x 3 126.1 x 3, 148.* x 3, 170.7 x 3 231.6 x 3, 258.4 x 3, 285.4 x 3 , 193.215.3 x , 03 237. 33 x x 6

Average He*t Flux (W/ca2) 31.7 31.9

Hot Spot Factor Nuclear Uncertainty Factor FH - FR'F2

Axial Pea o Averagt k 2 F e Fig. 19 Fig. 19

Radial Peak to Average FR 1.49 1.49 Uncertainty Factor for Bulk Water Temperaturb F e 1.30 1.30 Uncertainty Factor for Fll{ F a Tcapcrature 1.66 1.66

Coolant Velocity (n/aec) 3.80 4.03

Coolant Temperatur t Cora e e Inlet (*C) 57.1 57.1

Coolant Preaaure at Core Inlet (kg/cm A) 1.70 1.70 2 Bnr.Mutl" 2 c tub« pneu»« e ttubl « M-12 M u- KT-U HT-H (66*1

HT-10

HT-9

HT-«

tank

KT-7

O Vertical «*p«rl»cnt*l thl«61« Thermal column ildc © Control rod HT-6 6A 68 6CJ,cor« imaittlon hoi« 60. (C/Undrlcal fu«l «l*Mnt) KT-5 KT-* HT-3 Fig. 1 JRR-2 core configuration radial pattern « 2 Horizontal cross section of JRÄ-2 762

& tn O er> LO

95'

Cylindrical Type MTR Type

Fig, 3 JRR-2 Fuel Element

156 unit;cm , WTR Type Fuel Element , l l i 1 i i 1 ~~^ 0 —i CM Itu o 2: ^ U. ^ O ^^ U. ^J a ^£ au ^^ a I 0.038 —- y -— 3.7034 (~~

n .^n ^ 1 x- . v 0.0255 -j. >•_ 7^ • inm ininmin inininm ininininmJ o / into . t> Wi-fCM / T Or ^ o \ n i oooo^r«^ i . •~ —r « M C i cj •— imo^o» » o n v i o o o s / cMvo \ *- / rr O « O t M C M C O O .. O OC > M^"«^O O - r ...^ ...... / t O O O O O O OOOO •—t i—l i-H I-H «N Irtj

~~Extra Region O 90.9D. o 2v/ 9.08 v/oAl~~

~~Cylindrical Type Fuel Element~~

~~1 1 unit ; cm i i •Extra Region i03 o ,2 en r-t 0 2 D 87.5 o 9v/ U~~

~~I u» in Li n u•> i -n un /i •* t> ( •« "M C 1 ^K ( o *• D CO f** c\ 1 V0 C3 1/I Cn 1 C c l fro ^r v <*• ^T O CM~~

~~1 <=> 0 Ct 3 C5 C 3 C 3 3 CM 11 n U — —— it V*PeG Jl • 4. 1 1~~

~~Fig . 4 Models of JRR-2 MIR Type Fuel Element and Cylindrical Type Fuel Element used for Cross Section Preparation.~~

157 tO M U at en M oo MTR Type Fuel Elément •§ SJ l (n ni INNER J '0^ t-t W IUJ •< g D FUEL DO FUEL u. D20 2° + A. u (unit cell (sub cell o: la MODERATO! O MODERATOR m MODERATOR »-( averaged) c1 r- 1 averaged) 'u. < O < ëo 0.0255— « 0.24 — «H ^A. —0.52 0.63 — _«_ 1—— ————> N ^ *0.038 — u. -JÎ0.15--J ——— ^ _ 3.05 _^ ^_2.3948 ^J ___ 3.18 __ _ 2.3948 _j **0.076S ' **0.21 3* inne1 r plate l * oute* • r plate l 1 1 1 ^ __ IIMTT TRI U—1 ciin r-ni i —————— — SIIPCD rni i slab slab slab

~~Cylindrical Type Fuel Element 1 M t- 1 û~~

~~Fig.5 Detailed Models of JRR-2 MTR Type Fuel Element and Cylindrical Type Fuel Element used for Cross Section Preparation. Fig. 6 1/6 Core Model of JRR-2 used for FEDM-BURN Calculation.~~

~~159 1.20 „~~

~~Current Core 195g U-235/Element 93%Enr. MTR Type;20 Ele£ylindrical Typei4Ele< A measured values~~

~~1.15 -calculated values y FEDM-BURb N (MTR Type;18Ele. Cylindrical Type;6Ele.~~

~~1.10~~

~~0) ^~~

~~1.05~~

~~1.00 100 200 0 40 300 500 600 700 _ Burn-up (MWD)~~

~~Fig. 7 K eff vs Burn-up of JRR-2~~

~~160 2.0~~

~~1.5 HEU Core MEU Core LEU Core ; GE>183. V Ke 2 CE t— G,; E<0.6826 eV »—i en 1.0 ce~~

~~Xr> _i 0.5 u_~~

~~0.0 0.2 0.4 0.6 0.8 xlO~~

~~RRDIUS (CM)~~

~~Flg.6 JRR-2 Neutron Flux Distribution ( BOL ) 2.0 o\~~

~~1.5 * HEU Core MEU Core LEU Core~~

~~Œ j E>183.V Ke 2~~

~~CD 1.0 I ; E<0.6826 eV a: cr~~

~~x 0.5~~

~~0.0 0.2 0.4 0.6 0.8 xlO~~

~~RRDIUS (CM)~~

~~Fig.9 JRR-2 Neutron Flux Distribution ( EOL ) 1.20 Calculated value by FEDM-BURN~~

~~— HEU(93%Enr.)—° . 195gU-235/Element 18 MTR.6 Cylindrical Elements Core~~

~~t HEU(93%Enr.) 195gU-235/Element 24 Cylindrical Elements Core -TÏ— MEU(45%Enr.) 192gU-23S/Element 24 Cylindrical Elements Core 1.15 MEU(45%Enr.) 220gU-235/Element 24 Cylindrical Elements Core~~

~~LEU(20%Enr.) 237gU-235/Element Cylindrica4 2 l Elements Core~~

~~LEU(20%Enr.) 206gU-235/Element 24 Cylindrical Elements Core~~

~~1.10~~

~~<4-t~~

~~1.05~~

~~1.00 200 400 600 800 ————•- Burn-up (MWD)~~

~~Fig. 10 K eff vs Burn-up of JRR-2 Core~~

~~163 U Enrichment : 93% Fresh Fuel Loading : 19Sg U-235 MTR:18,Cylindrical:6 U-235,g(EOL) B°L Ke£f EOL Keff 1.0417 g , (EOLPu ) Bum-up 600 MWD~~

~~DistributioL EO 1 JRR-f Fig1 o U (93%.u 2P HE U-23f n o d )San Fuel~~

~~U Enrichment 45% Fresh Fuel Loading 220g U-235 Cylindrical:24 K L BO 1.1790 eff U-23S,g(EOL) EOL 1.0578 Pu,g (EOL) Burn-up 800 MWD~~

~~Fig. 12 EOL Distribution of U-235 and Pu of JRR-2 MEU (45%) Fuel~~

~~164 U Enrichmen% 20 : t Fresh Fuel Loading : 237g U-235 Cylindrical:4 2 199.S. \U-23S,g(EOL) KL efBO f: 1-170 1 1.714 /Pu ,g (EOL) EOL : 1.0575 Bum-uMW0 80 D : p~~

~~193.6~~

~~2.230 199.5~~

~~1.715 CR~~

~~DistributioL EO JRR- f 3 o 1 U (20%u 2P LE U-23f . no )d Fig Fuean 5 l JRR-2~~

~~1.6 -~~

~~inlec Fig. 14 Heac flux distribution in the JRR-2~~

~~165 JRR-2 JRR-2~~

~~Margin to ONB = 1.34 1.4- B Margi6ON o t n Margin to DNB - 11.6 11.- MargiB 9DN o t n~~

~~100 100 o u~~

~~OJ o0). s i" u 50 (U 50 H H~~

~~10 20 30 40 50 60 m c , z outlet inlet outlet inlet~~

5 Fue1 - l surfac d coolanan e t bulk temperature Fig.16 Fuel surface and coolant bulk temperature distribution JRR-e U th fue HE 2n li s distribution e JRR- th U fue n LE 2i sl element element APPENDIX E OF THE IAEA GUIDEBOOK ON CORE CONVERSION

~~ENRICHMENT REDUCTION CALCULATIONS FOR THE DR-3 REACTOR, RISOE, DENMARK~~

~~C.F. H0jerup~~

~~167 Introduction~~

This paper presents the calculations done on 9 actual reactor runs of a combined duration of about 200 full power days. All fuel elements were replaced at least once. A mixture of elements with either ~ 150 gr U235 or ~ 120 gr U235 initial loading were employed, and r comparison,fo , exactl e samth ye fuelling scheme was repeated with 20% enriched fuel. Three sets of initial loadings were tried, viz. (170, 136 gr U235), (160, 128 gr U235) and (164,131 gr. U235), the latter one being selected as the one where the EOC reactivity comes closest to the 93% case.

~~Calculation methods~~

Cell bur p calculationu n s were performed wit a programmh e (CGC, ref.l), which start t witou sa hcalculatio 0 grou1 f o pn microscopic cross sections for all isotopes. To do that the programme employs 76 group calculations with collision probabilities, thermalization calculations accordin a modifie o t g d Nelkin model, resonance calculation e detaileth n o s d annular geometry afte e sub-grouth r p method d condensatioan , 0 1 no t fro 6 7 m groups. r specifieFo d time step e programmth s e prepares tablef o s diffusion parameters, which are subsequently used by the overall diffusion theory programme DBU (réf. 2), which is able to interpolate its diffusion parameters from burn up tables.

5 energy groups were used in the diffusion calculations. The energy boundaries were 15 MeV - 2.865 MeV - 5.531 keV - 1.855 eV - 0.625 eV - 0 which means that the two upper groups from the benchmark study have been subdivided.

The DBU programme can be instructed via input data to replace fuel element y tim y an fresb et a sh o shufflonest r o , e elements (thi st use optiono d s here)nwa U deliverDB . e finath s l results, dux, power and burn up distributions.

~~168 Calculations performed~~

Cell bur p calculationu n s were performe a numbe r f fo fueo dr l % caso type93 loadingstw ef e elemento sth r Fo . s were present, the majority having an initial loading of about 150 gr U235 in 4 concentric fue l6 elementtubes4- d an , s havin e samth ge kind of fuel material, but only in the 3 outer tubes, which gives app. 120 gr U235 per element.

~~r elementg e powe0 Th 15 re s leveth 3 kW/c s taker 6. wa lfo o m t n r elementg 0 12 se th 0 whicankW/c 5. r d fo hm corresponde th o t s average power density in the core when operating at 10 MW.~~

This implies that the cross sections delivered for the over-all calculation contain the average Xenon poisoning, and no cor- rections were e madmesh-power e o th leveadjusX t e o e t lth t , allthough the DBU programme has an option for that.

~~e maiTh n objective with these calculations bein a comparisog n between a 93% case and a 20% case this inaccuracy in the Xe- treatmen s considerei t d insignificant.~~

A few cell calculations were made for the 93%, 150 gr case withou t wittbu h bur , varyinup n g concentration 0 1 addeB f o ds e heavtth o ye experimenta wateth f o r l space A .concentratio n of 2.2*10^-° B^ atoms per cc was found to reduce the reactivity d thian s , amounb4% y f B^-o t s the°wa n a constankep s a t t load in all the calculations

In the 20% case fuel elements with 170 gr U235 per element and 160 gr U235 per element were calculated and from them a value r U23g r elemen pe 54 s 16 e judgeappropriat b wa f t o o t d e from the poin f vieo tf obtainino w e samC reactivitth gEO e n i s a y the 93% case.

over-ale th r Fo l calculation e samth se modellin e reactoth f o gr as used in the benchmark was employed, with the exception that the full core had to be defined (instead of the quarter core), as the fuelling scheme is in no way symmetric.

169 A 'lay-out: of the core at the beginning of the calculations is give n figi n whic, .1 h show e distributioth s f U23o n 5 contents/ element in the 26 positions (these contents are estimated by the DR 3 staff's follow up programs).

~~It appears from fig .1 tha t fresh elements have been lodgen i d positions A2, C2, and E2, all with 4 fuel rings and approximately r U235/elementg 0 15 .~~

~~The reactor operation over the periods to follow was now reproduced quite closely as regards new fuel loadings and total energy production.~~

~~e operatioTh n schem s showi e e firs n Th tabli n. t1 e column holds the DR 3 period numbers. Then follow in the next columns:~~

~~2. The positions (referring to fig. 1) where fresh fuel element e beginnine addeth ar st a d f eaco g h period.~~

~~3. The type of element (150 gr or 120 gr element)~~

4. The equivalent number of days, which a fresh element with 150 gr U235 (120 gr U235) should have been irradiated in order to deplete its U235 contents to the value given by the manufacturer for the actual element.

~~. Irradiatio5 n tim n fuli e l power days.~~

Exactly the same scheme was afterwards applied to the 20% case. First, elements with initially 170 (136) gr U235 were used, and they gave consistently highe C reactivitiesEO r . Then elements with initiall 0 (12816 y U23r )g 5 were tried, which gave consistently lower EOC reactivities (all compared to the 93% case). s decideIwa t d that elements holding abou 4 (13116 t r U23g ) 5 of 20% enrichment would behave very much like the 150 (120) gr U23 C reactivit5EO element % s enrichmena 93 s r f wa o yfa s s a t concerned.

170 As verification of this statement seives fiq. 2 which shows ^eff vet"sus irradiation time; foi the two cases, (and the 170 qi r caseq 0 ans16 d indicate s well)a d . Some other results ftom e calculation th e variouth r Co e s, givear sk« n i tabl n, 2 e case s functioa s f burn-upo n ,% tabl93 e , mas3 eth f U23 o sn i 5 % case20 e san th dtogethe 3 staff'rDR wite th sh estimated values - which are based on measurements of the thermal flux

distribution -, and the keff values from the overall calculations n fig I e show. ar e ratio. th 3 n f thermao s l fluxef o s the 20%, 164 gr case to the 93%, 150 gr case at the beginning of the first and at the end of the last period calculated (i.e. with app. 200 full power days in between). Fig. 4 gives all the 5 group fluxes along the x-axis (refer to fig 1). In full lines are shown the 93%, 150 gr U235 case and as crosses the 20%, 164 gr case. Except for the thermal flux there is very little difference between corresponding fluxes.

~~Finally, in fig. 5, the mass of U235 distribution at the end of the las % castd compare 93 e perioan ee th th s giveo i dt r d fo n D3 Rstaff' s estimated values, partly base n measurementso d .~~

~~Conclusions~~

~~e calculationTh s presented here concernear e d exclusively with the reactor physical aspects of the enrichment reduction.~~

It can be seen from the results that the main negative consequence of reducing the enrichment to 20% is the reduction in therma e corel th flu n .i x Thi se ordereductioth f o f ro s i n 15% (down to 83 - 88% of the 93% case). The fast and epithermal fluxes are essentially unchanged for the same power level. The thermal flux in the reflector is also reduced, but to a smaller extent than in the core (to about 95% at the peak position).

~~e amounTh f U23o t 5 requirer q 4 r elemen16 pe d e s b founi t o t d of 20% enrichment to match the ROC reactivity of the presently used 150 qr, 93% elements.~~

171 The inactivity woi th of. tho con tin I aims will be sliqhtly L educed, but this pioblem has not been looked into yet. It seems faio t assumer , thouqh, l thateactivital t y incidents requit ing the action of the control system will be reduced by the same proportion, so that the system safety is essentially unchanged f I anything. e morth ,e negative fuel temperature coefficient, should act as an improvement of the safety with % thenriche20 e d fuel.

~~References~~

~~. C.F1 . Hojerup e ClusteTh . r Bur a Comp Programm u d n -an C CG e parison of its results with NPD-experiments. Risa-M-1898. 1976.~~

~~. K.E2 . Lindstram Jensen. Developmen d Verificatioan t f o n Nuclear Calculation Method r Light-Watefo s r Reactors. Rise-Repor . 235No t , 1970.~~

~~172 Table 1. Operational characteristics of the 93% cote~~

Period New elements Irr. t i me number Full Power Posit ion Type Days Days 234 A2 150 6.33 16.308 C2 150 3.80 E2 150 3.80 235 B3 150 3.59 23.125 B5 150 3.80 D2 120 0 D4 150 3.17 236 B4 150 3.59 23.188 C5 150 3.59 C6 120 0 El 150 3.80 237 A4 150 3.59 16.325 Bl 150 3.5-9 82 150 3.59 C4 150 3.38 238 C3 150 3.59 23.333 Dl 150 2.53 D5 150 2.11 239 Al 150 2.32 23 .333 A3 150 2.11 E3 150 2.32 E4 120 0 240 Cl 150 1.90 23.208 C2 120 0.19 D3 150 1.90 D6 150 1.90 241 B3 150 2.32 23.125 02 150 2.11 D4 150 2.32 E2 150 2.53 242 A2 150 2.74 30.208 Ek 150 4.22 B5 150 4.43 B6 120 0

~~173 Tabl . Infinit2 e e multiplicatio- en % 20 d an n % factor93 r fo s richments, 3 and 4 fuel tubes~~

Time 93% 20% Days 120 gr 15r g 0 131 gr 164 gr 0 1.717 1.766 1.639 1.672 0.5 1.665 1.711 1.590 1.621 1.5 1.644 1.690 1.571 1.601 10 1.622 1.668 1.552 1.581 20 1.608 1.654 1.540 1.570 40 1.583 1.630 1.519 1.550 80 1.526 1.577 1.474 1.507 120 1.456 1.511 1.420 1.456 160 1.363 1.424 1.356 1.395 200 1.235 1.302 1.278 1.321

~~174 . CompatTablmo i . . t 3 n Mas3 estaff' wit s iso f R o 1)25su O s h r vo 5s estimates k , ftom ovot-all calculations~~

~~Time _ Mass o£ U235, qrammes f f kp Full 93% 20% 93% 20% Power 3 estDR .j Cale . Calc. Calc. Calc. Days~~

~~0 2612 2612 3007 1.0302 1.0256 16.308 2408 2412 2815 1.0002 1.0041~~

~~0 2700 2702 3095 1.0510 1.0413 23.125 2411 2418 2822 1.0075 1.0095~~

~~0 2715 2723 3118 1.0517 1.0422 23.188 2425 2438 2845 1.0093 1.0113~~

~~0 2766 2780 3179 1.0563 1.0458 16.325 2562 2579 2986 1.0282 1.0253~~

~~0 2814 2829 3233 1.0617 1.0503 23.333 2522 2542 2958 1.0223 1.0216~~

~~0 2819 2840 3248 1.0568 1.0477 23.333 2527 2553 2973 1.0157 1.0176~~

~~0 2821 2840 3247 1.0566 1.0476 23.208 2531 2555 2973 1.0150 1.0170~~

~~0 2854 2867 3275 1.0650 1.0530 23.125 2565 2583 3002 1.0256 1.0242~~

~~0 2863 2881 3290 1.0640 1.0529 30.208 2485 2510 2934 1.0119 1.0149~~

~~175 l 2~~

~~1 2 3 4~~

~~105.6 147.0 120.8 73.3~~

~~90.5 82.9 74.6 92.1 76.2 108* 2 . B~~

~~122.0 148.2 127.2 101.7 78.5 73.2 +~~

~~D 84.0 76.2+ 134.3 75.4 87.2 + 106.9~~

~~75.3 148.2 110.7 92.3 +~~

~~fue3 + l U23. elemenr tubesgr pe 5 0 12 (t initially)~~

~~FIG. 1. Layout of core at start of calculations. Numbers are best estimates of grammes of U235 per fuel element~~

~~176 FIG. 2 . ——— 93%, 150 gr. 07 . veff versus time for various fuels. — — 20%, 164 gr. x---x . 20%gr 0 ,17 06 0--0 20%, 160 gr.~~

~~,05~~

~~.03~~

~~.01~~

~~.00~~

~~50 100 150 200 FULL POWER DAYS 2 3 4 5 1 2 3 4 0.865 0.874 0.872 0.867~~

~~0.864 0.865 0.853 0.857~~

~~0.858 0.84-1 0.838 0.850 0.854 0.901 B 0.865 0.845 0.852 0.856 0.861 0.885~~

~~0.863 0.861 0.855 0.850 0.852 0.882 0.870 0.860 0.840 0.837 0.840 0.867~~

~~0.856 0.853 0.858 0.842 0.869 0.884 0.871 0.862 0.853 0.857 0.853 0.879~~

~~0.852 0.871 0.868 ' 0.884 0.865 0.870 0.865 0.880~~

~~. 3 FIG. Thermal flux ratios. Ratio of 20%,164 gr. case to 93%,15 e beginnin th . cas gr 0t a e f firso g t period(upper numbers) and at the end of last period (lower numbers).~~

~~178 Graphite n O Core D 0 Graphite 2 î 1~~

~~FIG. 4. 5 group fluxes along x-axis. Full lines. : gr 93% 0 ,15 xxx : 20%, 164 gr.~~

~~Energy boundaries: - 2.86 V V 5Me Me 5 1 : 1 2: 2.865MeV- 5.531 keV 3: 5.531keV- 1.855 eV 4: 1.855 eV- 0.625 eV 5: 0.625 eV- 0.0001 eV~~

~~-100 x cm 100 2 3 4 5~~

~~1 2 3 4~~

~~100.6 130.9 97.9 86.4 100.8 130.2 96.3 87.0~~

~~86.0 76.3 114.4 127.5 129.7 106.8 B 88.4 74.0 113.1 126.0 129.3 108.2~~

~~110.0 79.9 73.7 67.3 65.5 60.6 109.4 78.2 69.5 63.6 63.1 61.6~~

~~92.4 117.1 100.7 114.6 85.7 114.2 92.9 115.8 98.6 112.3 84.0 112.6~~

~~76.5 118.7 97.0 79.4 77.0 118.1 95.5 80.6~~

FIG. 5 . e th Distributiof o d en e th f U23n o t a 5 last calculated reactore perioth r fo d 93% case. Comparison with DR3' n estimatesow s based on flux measurements. Upper calculatee numberth e ar s d values, lower measuree numberth e ar sd values.

~~180 APPENDIX F~~

~~Benchmark Calculations~~

~~F-0 Specifications ...... 182~~

~~The benchmark calculations were performed by the following organizations:~~

~~(USAL AN F-)l ...... 5 18 .~~

~~F-2 HARWELL (UK) ...... 201 F-3 AAEC (Australia) ...... 5 21 .~~

~~F-4 JAERI (Japan) ...... 231 F-5 RIS0 (Denmark) ...... 239~~

~~ABSTRACT~~

Benchmark calculation neutronicf so safety-related san d parameters were performed to compare the computational method variouf o s s organizations - methode re Th *d an s sults for an idealized core are described in Appendices F-F-5o t l . Only limited conclusion actuar fo s l conversions from HEU to LEU fuel should be drawn from these results.

~~181 APPENDIX F-0~~

~~Specifications for Methodical Benchmark Probleo Heavr fo y Water Reactors~~

~~Aims; Compariso differenf no t calculatlonal method cross-sectiod san n data sets used la different laboratories» Limited conclusions for real conversion problems» Reactor Design Specifications~~

Total Power 10 MWth (power buildup by 3.1 x 10*° fissions/Joule) Active Core Height 610 ma distancm m Extrapolatio7 22 en fro(i coree m e mth m ,th n7 Lengt22 h cosine-shaped flux coree goe th zero t s,n o i heav y water reflector, and graphite reflector regions) Lattice Pitch 152 mm Aluminum Density 2.7 g/cm3 Graphite Density 1.7 g/cm3 Heavy Water Temperature 20"C Fuel Temperature 20*C Xenon-State Equilibrium Xenon

~~Heavy Water Composition 99.7 DZ 25wt 0 ,° 0.2%2 Z 5wt~~

Specifications of the Different Fuels (UA1X-A1) for HEU, HEU, and LEU 235o Uw/ 3 9 HED: Enrichment 2350 per fresh element 150 g voido v/ )7 wt( ZD 18.4 U Density 0.547 g/cm3 MEU*: Enrichment 45 w/o 23SU 235fresr upe h element 156.3g wtZ U (7 v/o void) 33.7 Densit0 y 1.177 g/cm3 LEO: Enrichment 20 w/o 235U 235g per fresh element 167.5 g wtZ U (7 v/o void) 58.3 0 Density 2.838 g/cm3

~~Results Heutronies burnu. celvs r pfo lk calculation • s w/owit0 2 hd . enrichmentan o w/ 3 9 f so~~

« Reaction rate diffusiod san n coefficient thren si e energy groups from cell calculation zerr sfo o 235U burnup witw/o0 2 hd . enrichmentan o w/ 3 9 f so for BOC and EOC corea «ich enrichments of 93 w/o and 20 w/o. « corFluxee th e t midplansa e alon x-axie gth thren si e energy groups for 93 w/o core at BOC. • corFlue caseo eth x w/ midplant ratiosa 3 9 d s an e betweeo w/ 0 n2 along the x-axis in three energy groups at BOC.

~~Energy Group Reportinr sfo g Results~~

~~«thermal 0 eV < Ea < 0.625 eV 5.53< » E 10.62ke < V 5e 'epithermal n *fast~~

~~^Calculations with MEU fuel are optional.~~

~~182 Specifications for Safety-Related Calculations~~

r computatioFo f safety-relateo n coreso e w/ ar 0 2 d dan parameterso w/ 3 9 e th , uniformly loaded with fuel elementU n whic23S havI g s 0 e4 h been burned. Computed concentrations of 236U, 238U, various plutonia, and fission products are to be Included. The densities of heavy water at 20"C and 50°C are 1.105 and 1.096, respectively. The densitie f ligho s t wate t 20*a 50°r d e 0.99 Can C0.988ar d 8an , respectively.

Results - Safety-Related • Reactivity change due to uniformly increasing the temperature of the materials In all coolant channels and all fuel plates from 20°C to 50°C. The moderator temperatur s 20°Ci e . Ignore axial expansion.

• Reactivity change due to uniformly increasing the temperature of the materials in all fuel plates from 20°C to 300*C. Coolant and moderator temperatures are 20°C. Ignore axial expansion. • Reactivity change due to decreasing the coolant density in the two central fuel elements by 20Z. Coolant and moderator temperatures ore 20°C. The extrapolation length is 227 mm.

~~• Optiona - Prompl t neutron generation delayetimd an e d neutron fraction.~~

Optional Reaction Rate Ratios Core: 40 g 235U uniform burnup BOG core Isotopes: 235U, 98Mo, 10B, Co, Si (infinite dilution cross sections in 020 inside central fuel tube) Locations: Core center and 76 mo into reflector along x-axls (532 mm from core center). Results: Total absorption reaction rate ratiod s an betweeo w/ 0 n2 93 w/o cases for each Isotope.

~~Fuel Element Geometry Specifications~~

~~Geometry Data~~

~~Component Thickness,~~

~~Fuel Tubes*(4) 1.52 Outer Al Tube 1.57 Coolant Channels 3.38 Fuel Meat 0.50 Clad 0.51~~

*Fuelea dm Lengt0 61 - h

~~Cell Volume Fractions~~

~~Fuel Meat 0.0209 0.9147 \ 0.0644~~

~~183 Methodical Benchmark, 10 MW Heavy Water Reactor CorModelY X eC CrossEO d san SectionG BO r sfo~~

~~Outside Boundary Condition t • 0~~

~~t 2 Fuel Cell Width Graphite Reflector BOC Core~~

~~Fue3 l Cell Width Heavy Water Reflector O 99.7D. o 5w/~~

~~0 2 H 0.2o w/ 5~~

~~20 30 Crams U Burned~~

~~t Fue3 l Cell Width 2 Fuel Cell Outside 152 ma 20 30 _ 99.75 w/o D20 Width _^ Boundary Graphite Condition ; _ _L H,,0.2o w/ 50 Reflector * - 0 22.5 22.5 10~~

~~m m 2 15~~

~~235 CramBurneU s d EOC Core 30~~

~~60 30 40 Grain burneU s d means losU fissioy tb d nan capture. 30 30 15~~

~~Grans 235U burned per >j element~~

~~184 F-l.l~~

~~APPENDIX F-l~~

~~Benchmark Calculations for Heavy Water Reactors~~

~~J. E. Matos and K. E. Freese~~

~~Reduced Enrichment Research and Test Reactor~~

~~(RERTR) Program~~

~~Argonne National Laboratory Argonne, Illinois 60439~~

~~U.S.A.~~

~~185 F-1.2~~

~~CALCULATIONA R SFO BENCHMARK HEAVY WATER REACTOR USING DIDO-TYPE FUEL ELEMENTS WITH HIGH, MEDIUMENRICHMENTW LO D ,AN S~~

Introduction A benchmark problem for a heavy water reactor has been calculated in orde comparo t r e calculational methods use varioun i d s research centerr fo s both neutronics and safety-related parameters. Detailed specifications for the benchmark proble showe mar Appendin i n x F-0. They, corresponMW 0 1 a o t d 26 element core at beginning (BOG) and end (EOC) of equilibrium cycle. The cor surroundes ei innen a y rb d reflecto heavf ro y outen watea d rr an reflecto r of graphite. Uranium enrichments of 93%, 45%, and 20% are considered in DIDO-type cylindrical fuel elements containing fresh 235u loading 150.0f so , 156.3, and 167.5 grams, respectively. Diffusion theory calculations are required in X-Y geometry only.

Sinc purpose eth f thio e s benchmark proble comparo t s i m e calculational methods, the reactor configurations were Idealized and simplified. Thus, these calculations may not correspond to realistic reactor conditions, and only limited conclusions about actual reactor performancU ME d e an wit U hLE fuels should be drawn from them, even though some results are very similar to the results of the specific studies.

Neutronics Calculations Cross Section Burnur sfo p Calculations Five-group microscopic cross sectioncore th e r wersfo e generates da a function of burnup using the EPRI-CELL code with the methods described for the benchmark calculation IAEe th A n Guidebooi s Researcn ko h Reactor Core Conversio Highlf o e nyUs frow EnricheEnrichee Lo th mf o de d Us Uraniue th o mt Uranium Fuels (for Light Water Reactors, IAEA-TECDOC-233, August 1980)e Th . five energy group structur e calculation th use n three i d th ed grousan p structure used for presentation of results are shown in Table 1. The fuel element was modeled as shown in the specifications with all four fuel tubes outee anth d r aluminum tube explicitly defined e thicknesTh .e rin f th o g f so heavy water outside the fuel element was chosen to preserve the volume of heavy water associated with each element simplcityr Fo . cel e s chose,th lwa n to be infinitely high. The heavy water was assumed to contain a light water impurit 0.25%f yo . In order to validate the microscopic cross sections and isotopic reaction rates, Monte Carlo calculations were performed for zero 235u burnup using the identical cell geometry as for the EPRI-CELL calculations. Comparisons of the EPRI-CEL Montd Lan e Carl o235e valueUth r fissiosfo n cross sectione th d san 238jj absorption cross sections for the 93%, 45%, and 20% enrichment cases are shown in Table 2. Isotopic reaction rates normalized to one neutron lost are shown in Table 3 for the 93% and 20% cases. The five-group structure used in the calculations was collapsed to the standard three-group structure defined i ne agreemen TablTh . e1 t betwee EPRI-CELe nth Montd Lan e Carlo dats i a considere vere b yo t dgood .

~~186 F-1.3~~

~~Atom densitie fuee EPRI-CELe th l th n measi r fo t L calculation 93%e th ,r sfo % enriche20 d 45%an ,d fuel element variout sa s stage burnuf so showe par n i n Tables 4, 5 and 6, respectively«~~

Microscopic cross sections in three energy groups for 235y an(j 238u as a functio gramf no s 23SU burne presente e l threar dal er Tablen fo enrichi d 9 s7- - ments. Separate cross sections were also prepared for the reflectors using a homogenized core sourc d appropriatan e e thicknesse f heavso y wate d graphitean r . The resulting diffusion coefficients and microscopic absorption cross sections heave foth r y wate d graphitran e reflector showe sar Tabln i n. e10

Diffusion coefficients froEPRI-CELe th m L calculatio corr nfo e cross sections at zero 235U burnup are shown in Table 11 for the 93% and 20% cases. Tabl present2 e 1 compute^ k e sth d usin EPRI-CELe gth L functiocoda s ea n of grams 235U threl burneal er fuefo d l enrichments.

Result Reactor fo s r Calculations o dimensionaTw l diffusion theory calculations were performe Y geometrX n i d y usin DIF2e th gspecifiee D th cod d ean d reactor configuration A 1 tota.4 f o l mesh intervals in the x-direction and 33 in the y-direction was used, including 5 mesh points per fuel element and 15 mesh points in the heavy water reflector. e fluxeTh s quartee werth en i normalizer W coreM 5 powea , 2. o witt f dro n a h axial half height of 305 mm and an extrapolation length of 227 mm. Values of keff for the 93% enriched core with all fresh fuel elements, as well as the coreC 93%r EO fo sd BO % ,enrichmentsan C 45%20 d ,an showe ,ar Tabln ni . 13 e

Fast, epithermal, and thermal fluxes for a midplane traverse along the x-axis % througenriche93 core e centee showe th th th h ear dFign f r i no r fo .1 BO Caverage th case e pead Th e.an k thermal fluxe core sth e (<0.62n i ) 5eV are computed to be 1.55 x 10^ and 1.36 x 10^ n/cm^/s, respectively. In the heavy water reflector e pea,th k thermal flu s 1.3xi 110^-x ^ n/cm^/a t sa distance of 70 mm from the active core. Fast, epithermal d therma,an fluC % enrichelBO x45 e ratioth df so cas e % enriche% enriche20 93 % e enrichee 93 th tth o de df o th cascas dd o t eean cas e are respectively, show3 Figsd n ni an .2 midplan a r ,fo e traverse alone th g x-axis through the center of the core. The fast and epithermal fluxes for both % enrichment20 d an essentialle % thar 45 es y unchange r greateo d r than thosf o e the 93% enriched reference core. The peak and the average thermal flux ratios in the core for the 45%/93% case are about 0.95, while the thermal flux ratio at the peak in the heavy water reflector is 0.98. For the 20%/93% case, the peak and the average thermal flux ratios in the core are 0.86 and 0.88, respectively. The thermal flux ratio at the peak (also at 70 mm from the active core) in the heavy water reflector is 0.95. Safety-Related Calculations Specifications For computatio safety-relatef no d parameter % enriche20 d san wit d% h93 fuels reactoe ,th r specifications were changed such tha core s uniformlth tewa y loaded with fuel elements in which 40 g U had been burned. Computed concentrations of U, U, various plutonia 235, and fission products were included. 236 238

~~187 F-1.4~~

Cross Sections Ten-group microscopic cross sections (see energTablr fo 1 ey boundaries) for the core were generated as a function of burnup using the EPRI-CELL code in orde provido t r e greater e thermadetaith n li l energy range e singlTh . e thermal group (<0.625 eV) of the five-group structure is not sufficient to provide accurate results for reactivity changes due to spectral hardening with increasing coolant temperature.

Prompt Neutron Generation TimDelayed ean d Neutron Fraction Two-dimensional diffusion theory problems werusinn e ten-grouru e th g p EPRI-CELL cross section yielo t s d both rea adjoind lan t fluxes werd ,an e followed by diffusion-perturbation-theory calculation delayef so d neutron fractiod nan prompt neutron generation time. Table 14 gives the results for keff, &eff, and e correspondinTh . A g decay constant d delayesan d neutron fraction family b s y are shown in Table 15. Isothermal Reactivity Feedback Coefficients Reactivity changes were specified for three cases: Uniform increase of temperature from 20°C to 50°C in all coolant channels and fuel plates, including attendant decreas coolann i e t density. - Uniform increase of temperature from 20°C to 300°C in all fuel plates. - Uniform decrease of coolant density by 20% in the central two fuel elements. resulte specifiee Th th f so d calculation % enriche20 d san witd % fuelh93 s are e coolanshowTh Tabln ni defines . heavl i te16 al ys a dwate d lighan r t water inside the outer aluminum flow tube of each fuel element. Since the lowest temperatur deuteriur efo m L cross-sectioavailablAN e th n i e n library is 25°C, the moderator temperature was fixed at 25°C in all cases. However, calculations were performed for several coolant temperatures between 25°C and 60°C, and the results extrapolated to 20°C to obtain the values reported in moderatoe Th Tabl . e16 r tempertur extrapolatet no s ewa d since only reactivity difference importane ar s t here.

Additional calculations (Table 17) were also done for the case with uniform temperature increas coolane th n ei t channel fued an sl plate orden i s separo t r - ate the contributions to the reactivity change due to three physical effects. These effect hardenine th neutroe s) th are (1 f o g:n spectru increasino t e du m g the coolant temperature increase th ) neutron (2 ,ei decrease nth leakago t ee edu in coolant density, and (3) the increase in neutron absorption due to the Doppler effect in the fuel meat. Because of the harder spectrum with the 20% enriched fuel, the temperature change component is slightly smaller and the density change component is slightly larger than for the 93% enriched case. However, since the Doppler component is much larger with the 20% enriched fuel, the overall reactivity chang abous i e % largert8 . It should be noted that the extrapolation length was fixed at 227 mm in benchmarl al k calculation purposer fo s f simplicityo s actualityn I . e ,th extrapolation lengths change with enrichment and particularly with the density of the coolant. Thus, the results in Table 16 and 17 involving coolant density changes are viewed as indicative only. In calculations for specific reactors, changes in extrapolation lengths will need to be accounted for.

~~188 F-1.5~~

~~Reaction Rate Ratios~~

The purpose of these calculations Is to exhibit the difference, if any, f thermao betweee f reactious o l e e fluth nus xe n ratioth rat d ean s ratios to compare reactor performanc % enriche20 d ean witd % fuelsh93 . Total absorption reaction rates, including (n,Y), (n,f), (n,o), (n,p), (n,2n), etc., and reaction rate ratios were computed for the specified isotopes at infinite dilution. Sinc reactione th mos f to s othe(n,fd an r) )y tha, hav(n n e thresh- holds of several MeV or greater, only the (n,y) and (n,f) reactions provided significant contribution e totath o lt sreactio n rates. Cross Sections Infinite dilution cross sections were specifie simplicitr fo d y since the geometries of the targets and target-holders that are used for isotope production vary from reactor to reactor. To obtain cross sections appropriate for the center of the core, the EPRI-CELL calculation that was used for the uniforg 0 4 m burnup cors repeatewa e d with isotopic concentrationf so 10~x 0 1. at/bn*cra homogenize heave th yn di wate r insid centrae eth l 20fuel tube. The fine-group microscopic cross sections were collapsed into broad groups using fluxes spatially averaged over the central heavy water region only separatA . e cell calculation usin homogenizega d core source an larga d e regio heavf no y wate s repeaterwa obtaio t d n isotopic cross sections appropriate for the reflector region. Fluxes specifiee Fluxeth t sa d location reference th n si e whole-core calculation already performe computatior fo d f safety-relateo n d parameters were used along with cross sections frocele th ml calculation computo st e th e isotopic reaction rates whole-corA . e uniforg calculatio 0 4 e m th burnf no - up core was not repeated since flux perturbations caused by including isotopes with concentration 10~f so 2" at/bn«cr significante ab wil t lno . Results resulte Th thesf so e calculation showe ar sTabln ni . Absolute18 e reactio nthree-groue rateth n i s p structur faste th e show ,ear r epithermalfo n , and thermal energy ranges since this breakdown provides additional information for comparisons among the various Isotopes and as a function of energy for individual isotopes. Comparison of the integrated reaction rate ratios with the thermal «0.62 flu) 5eV x core ratioreflectoe th th e t n centea si d r an rsho w thae th t thermal flux ratio is a good indicator of performance for those isotopes that are approximately 1/v in the thermal range and do not have significant epithermal absorption. Among the specified isotopes, Mo is an exception since totae 75-80th lf % o absorption s epitherma e occuth n ri l range. 98

~~% enriche93 e Foth r d case (n,2ne th , ) reaction rat Groun s i ewa 1 p computed to be 2.13 x 10~9 in 235U and 1.51 x 10"12 in Si.~~

~~189 F-1.6~~

Table 1. Energy Group Structures Five-Group Structure Three-Group Structure Used for Neutronics Calculations Used for Reporting Results eV E ,eV E ,eV Group V L Group VeV L 1 1.0 x 107 s 10 8.2x 1 1 1.0 x 107 3 10 5.5 x 3

~~2 8.21 x 105 3 10 5.5x 3 3 3 10 5.5 x 3 1.855 2 3 5.510 3x 0.625 4 1.855 0.625 5 0.625 2.53 x 10-* 3 0.625 2.5 10-x 3 1~~

Ten-Group Structure Used for Safety-Related Calculations eV E ,eV „ E..,eV E ,eV Group V L * U Group L 1 1.0 x 107 6.39 x 10s 6 1.166 0.625 2 s 10 6.3 x 9 3 10 9.1 x 2 7 0.625 0.417 3 3 10 9.1 x 2 3 10 5.5 x 3 8 0.4170.146 4 3 10 5.5 x 3 1.855 9 0.1460.057 5 1.855 1.166 10 0.057 2.53 x 10-^

~~190 F-1.7~~

~~Table 2. Comparisons of EPRI-CELL and Monte Carlo Cell U Fission and 238U Absorption Cross Sections wit h235 93%, Enrichment% 20 45%d an , Zert sa o 235U Burnup~~

93% Enrichment 235U Fission, b 238U Absorptionb , Group EPRI-CELL Monte Carlo (±lq.%) EPRI-CELL Monte Carlo (±la.%) 1 1.779 2.181 (±0.37) 0.392 0.497 (±0.43) 2 26.698 26.402 (±0.53) 27.730 25.694 (±2.90) 3 331.769 328.820 (±0.26) 1.620 1.607 (±0.24)

~~45% Enrichment 23SU Fission,b 238U Absorption,b~~

Group EPRI-CELL Monte Carlo (±lqt%) EPRI-CELL Monte Carlo (±lq.%) 1 1.779 2.189 (±0.38) 0.392 0.497 (±0.56) 2 26.442 26.055 (±0.56) 13.614 13.932 (±1.85) 3 327.512 323.020 (±0.29) 1.601 1.581 (±0.29)

20% Enrichment 235U Fission,b 238U Absorption, b Group EPRI-CELL Monte Carlo (±lo.%) EPRI-CELL Monte Carlo (±lo.%) 1 1.779 2.186 (±0.37) 0.392 0.495 (±0.34) 2 26.173 25.719 (±0.53) 7.626 7.645 (±2.17) 3 319.940 316.950 (±0.37) 1.567 1.554 (±0.36) v«

~~Enrichment,% EPRI-CELL Monte Carlo (±lqt %) 93 1.8831 1.8936 (±0.48) 45 1.8382 1.8346 (±0.43) 20 1.7827 1.7780 (±0.52)~~

~~191 F-1.8~~

Table 3. CoaparUon of EPRI-CEU. and Monte Cerlo laotoplc Reaction Ratee with 93X and 20X Enrlchmenta at Zero "Sy Burnup Normalized to One Neutron Loat*

~~93X Enrlctaent~~

~~235u v-Fleelon 235(1 Flaelon 235u Capture Dauccrlum Capture Group EPRI-CELL Honte Carlo (Ha) X , EPRI-CELL Mont« Carlo (Ha, X) EPRI-CELL Monte Carlo (Ha, X) EPRI-CELL Honte Cerlo (Ho) X ,~~

1 4.3675-3 5.5293-3 (tO.42) 1.7579-3 2.2021-3 (10.42) 5.0565-4 5.6261-4 (10.55) 6.5189-6 5.0042-6 (±0.56) 2 9.5094-2 9.5178-2 (10.57) 3.9314-2 3.9349-2 (±0.57) 1.9964-2 2.0141-2 (10.77) 7.5800-5 7.6144-5 (10.27) 3 1.7762+0 1.7929+0 (10.51) 7.3433-1 7.4122-1 (tO.Sl) 1.2632-1 1.2757-1 (10.50) 1.9271-3 1.9485-3 (±0.4») Total 1.8757+0 1.8936+0 (10.48) 7.7540-1 7.8277-1 (±0.48) 1.4679-1 1.4827-1 (10.45) 2.0095-3 2.0296-3 (10.48)

~~238« v-Fleeloa 238U Fleelon !38u Capture Hydrogen Capture Croup EPRI-CELL Honte Carlo (tlo) X , EPRI-CELL Honte Carlo (Ha, X) EPRI-CEU, Honte Carlo (Ho) X , EPRI-CELL Monte Carlo (Ho) X ,~~

1 2.0396-5 3.6680-5 (10.97) 7.2510-4 1.3026-5 (10.93) 2.1531-5 2.4281-5 (10.61) 9.7150-7 1.0097-6 (10.30) 2 1.4939-8 1.4*13-8 (114.60) 6.4404-9 6.2133-9 (±14.60) 3.0349-3 2.8460-3 (12.90) 1.2212-4 1.2365-4 (10.28) 3 0.0 0.0 0.0 0.0 2.6656-4 2.6929-4 (±0.50) 3.1035-3 3.1421-3 (±0.49) Total 2.0411-5 3.6695-5 (10.97) 7.2574-4 1.3032-5 (10.93) 3.3230-3 3.1396-3 (12.63) 3.2266-3 3.2668-3 (10.48)

~~AltMlnua Capture Oxygen Capture Total (n. 2n) Croup EPRI-CELL Monte Carlo (tlo, X) EPRI-CELL Monte Carlo (Ho) X , leotope EPRI-CELL Hont« Carlo (Ho, X)~~

1 7.7543-4 1.0399-3 (±1.46) 2.7757-3 3.3153-3 (12.23) 2JSU 2.4383-6 4.6146-6 (±4.23) 2 2.4572-3 2.9612-3 (±0.41) 1.6069-5 1.2130-5 (±0.31) 238U 3.5242-7 7.1757-7 (±5.90) 3 5.3792-2 6.0178-2 (±0.49) 3.3396-4 3.3454-4 (10.49) Deuterium 3.6886-3 3.7370-3 (12.22) Total 6.2425-2 6.4180-2 (±0.46) 3.1257-3 3.6619-3 (12.02) Total 3.6914-3 3.7423-3 (12.22)

~~20X EnrlchMac~~

~~23Sy v-Flaelon 235|) riaelon 23!U Capture Deuterium Capture Group BPM-CELL Monte Carlo (Ho, X) EPRI-CELL Monte Cerlo (Ho, X) EPRI-CELL Honte Carlo (Ho, X) EPRI-CELL Monte Carlo (Ho, X)~~

1 4.8702-3 6.1913-3 (±0.39) 1.9603-3 2.4665-3 (10.39) 5.6393-4 6.3221-4 (10.51) 6.5069-6 5.0544-6 (±0.55) 2 1.0129-1 1.0077-1 (10.52) 4.1876-2 4.1659-2 (10.52) 2.1132-2 2.1179-2 (10.78) 7.2184-5 7.2409-5 (±0.36) 3 1.6680+0 1.6689+0 (±0.55) 6.8958-1 6.8997-1 (10.55) 1.1881-1 1.1894-0 1(1 .55) 1.6612-3 1.6624-3 (±0.49) 1.7399-3 (±0.47) Total 1.7741+0 1.7759+O (±0.52) 7.3342-1 7.3410-1 (±0.52) 1.4051-1 1.4075-1 (10.47) 1.7399-3

i33 238u v-Fleelon 238» Fleelon 4 U Capture Hydrogen Capture Group EPRI-CELL Honte Carlo (±lo) ,X EPRI-CELL Honte Carlo (Ho, X) EPRI-CELL Monte Carlo (Ho) X , EPRI-CELL Monte Carlo (Ho, X) 9.70U-7 1.0089-6 (10.25) 1.2073-3 2.1619-3 (10.87) 4.2905-4 7.6775-4 (10.35) 1.2760-3 1.4388-3 (±0.51) 1.1628-4 1.1759-4 (±0.35) 2 8.7630-7 8.1778-7 (114.50) 3.7777-7 3.5255-7 (114.60) 4.8188-2 4.8904-2 (12.15) .54) 2.6752-3 2.6808-3 (10.49) 3 0.0 0.0 0.0 0.0 1.3337-2 1.3362-2 (10 2.7925-3 2.7994-3 (±0.47) Total 1.2082-3 2.1627-3 (10.87) 4.2943-4 7.6810-4 (10.85) 6.2802-2 6.3705-2 (H .63)

~~) Tota2n , (n l Aluminum Captura Oxygen Capture EPRI-CELL Monte Carlo (Ha) X , Group EPRI-CELL Monte Carlo (Ho) X , EPRI-CELL Monte Carlo (Ho, X) leotope~~

23SU 2.7166-6 5.2121-6 (±3.64) 7.4647-4 9.8094-4 (±1.17) 2.7686-3 3.3571-3 (12.08) 2.0869-5 2 2.6239-3 2.6989-3 (±0.49) 1.5385-5 1.1511-5 (10.38) DeuterluaL 3.6794-3 3.7870-3 (±2.39) 3 4.8171-2 4.8996-2 (±0.50) 2.8790-4 2.8544-4 (10.49) 3.7030-3 3.8328-3 (12.36) Total 5.1541-2 5.2676-2 (10.47) 3.0719-3 3.6541-3 (H.9I) Total

~~192 F-1.9~~

~~Table 4. Atom Densities In 93* Enriched Fuel Meat vs. 23SU Burned Atom Densities (atons/cn 10* 3 2"*)~~

23S U Burned, g l«Xe ll»9S» 2350 23 Su 238(j 0.0 0.0 0.0 1.302-3 0.0 9.679-5 20.0 1.299-8 9.410-8 1.129-3 2.761-5 9.611-5 30.0 1.192-8 8.629-8 1.042-3 4.123-5 9.575-5 40.0 1.104-8 957-. 7 8 9.551-4 5.463-5 9.540-5 45.0 1.067-8 7.668-8 9.117-4 6.131-5 9.521-5 60.0 9.404-9 6.742-8 7.815-4 8.110-5 9.465-5

235 2 U Burned, g 239Pu *°Pu 2mPu 2*2Pu Al 0.0 0.0 0.0 0.0 0.0 5.399-2 20.0 5.875-7 2. 958-8 2.831-9 6.512-11 5.399-2 30.0 8.147-7 6.260-8 8.846-9 3.296-10 5.399-2 40.0 9.986-7 1.041-7 1.927-8 1.032-9 5.399-2 45.0 1.078-6 1.276-7 2.636-8 1.653-9 5.399-2 60.0 1.260-6 2.057-7 5.517-8 5.213-9 5.399-2

~~Table 5. Atoa Densitie EnricheZ 45 s n i d Fuel Meat vs. 2350 Burned~~

~~Atom Densities (a tons /cm3 x 102"*)~~

23SU Burnedg , l»Xe l-»9Sm 235u 23 6u 238u 0.0 0.0 0.0 1.357-3 0.0 1.638-3 20.0 1.364-8 9.916-8 1.184-3 2.769-5 1.632-3 30.0 1.264-8 9.173-8 1.097-3 4.141-5 1.628-3 40.0 1.180-8 8.534-8 1.010-3 5.480-5 1.625-3 45.0 1.143-8 8.253-8 9.664-4 6.158-5 1.623-3 60.0 1.021-8 7.344-8 8.362-4 8.141-5 1.619-3

~~2 2g Burned3S g , 239Pu 2M)pu 2-Hpu 2-» Pu Al~~

0.0 0.0 0.0 0.0 0.0 5.162-2 20.0 5.374-6 2.610-7 2.474-8 5.449-10 5.162-2 30.0 7.524-6 5.559-7 7.792-8 2.765-9 5.162-2 40.0 9.307-6 9.269-7 1.699-7 8.623-9 5.162-2 45.0 1.010-5 1.142-6 2.339-7 1.387-8 5.162-2 60.0 1.198-5 1.856-6 4.931-7 4.401-8 5.162-2

~~193 F-1.10~~

~~Tabl . 6 eAto m Densities.i EnricheZ n20 d Fuel Mea . 235vs t U Burned Atom Densities (atoms/cm3 x 102"*)~~

235 135 U Burned, g Xe 23S0 236U 0.0 0.0 0.0 1.454-3 0.0 5.744-3 20.0 1.479-8 1.079-7 1.281-3 2.791-5 5.730-3 30.0 1.385-8 1.011-7 1.194-3 4.177-5 5.723-3 40.0 1.307-8 9.506-8 1.107-3 5.538-5 5.716-3 43.0 1.273-8 9.242-8 1.064-3 6.216-5 5.711-3 60.0 1.157-8 8.373-8 9.334-4 8.212-5 5.702-3

23S U Burned, g 239PU 2*2Pu Al 0.0 0.0 0.0 0.0 0.0 4.538-2 20.0 1.189-5 5.473-7 5.131-8 1.045-9 4.538-2 30.0 1.680-5 1.170-6 1.625-7 5.333-9 4.538-2 40.0 2.102-5 1.965-6 3.581-7 1.673-8 4.538-2 45.0 2.291-5 2.421-6 4.920-7 2. 676-8 4.538-2 60.0 2.761-5 3.963-6 1.045-6 8.418-8 4.538-2

~~Table 7. EPRI-CELL Cross Sections vs. 235U Burned for 932 Enrichment 235p 238u~~

~~235U Burned ,g Group ffa °f 9 "f 0.0 1 2.291 1.779 0.392 9.887-2 2 40.255 26.698 27.730 5.885-5 3 388.840 331.769 1.620 0.0~~

~~20.0 1 2.291 1.779 0.392 9.886-2 2 40.468 26.820 27.767 5.875-5 3 397.020 338.790 1.651 0.0~~

~~30.0 1 2.291 1.779 0.392 9.886-2 2 39.588 26.881 27.785 5.871-5 3 405.040 345.670 1.683 0.0~~

~~40.0 1 2.291 1.780 0.392 9.886-2 2 40.693 26.942 27.803 5.866-5 3 413.370 352.830 1.716 0.0~~

~~45.0 1 2.291 1.780 0.392 9.885-2 2 40.756 26.973 27.812 5.864-5 3 417.700 356.550 1.732 0.0~~

~~60.0 1 2.291 1.780 0.392 9.885-2 2 40.394 27.070 27.839 5.857-5 3 431.530 368.420 1.786 0.0~~

~~194~~

~~F-l.ll. Tabl . 8 eEPRI-CEL L Cross SectionU Enrichmen1 235 . Burne45 «v» r fo d t~~

~~23SP Burnedg , Croup "a "f \ •t 0.0 1 2.291 1.779 0.392 9.883-2 2 39.818 26.442 . 13.614 5.887-5 3 383.882 327.512 1.601 0.0~~

~~20.0 1 2.291 1.779 0.392 9.882-2 2 40.028 26.555 13.650 5.878-5 3 390.700 333.360 1.627 0.0~~

~~30.0 1 2.291 1.779 0.392 9.882-2 2 40.128 26.608 13.668 5.874-5 3 398.050 339.670 1.65691 0.0~~

~~40.0 1 2.291 1.780 0.392 9.882-2 2 40.227 26.659 13.687 5.870-5 3 405.640 346.190 1.685 0.0~~

~~45.0 1 2.291 1.780 0.392 9.881-2 2 40.276 26.685 13.697 5.868-5 3 409.660 349.640 1.701 0.0~~

~~60.0 1 2.291 1.780 0.392 9.881-2 2 39.428 26.769 13.726 5.863-5~~

~~3 422.370 360.580 1.750 0.0 Tabl . 9 eEPRI-CEL L Cro*a Section* vs.'U Enrichmen Z 235 Burne20 r fo d t _____ 23S„ 238,,~~

a 235U Burnedg , Group 0 "f a °f 0.0 1 2.291 1.779 0.392 9.873-2 2 39.381 26.173 7.626 5.978-5 3 375.064 319.940 1.567 0.0 20.0 1 2.291 1.779 0.392 9.872-2 2 39.575 26.273 7.640 5.971-5 3 380.010 325.170 1.586 0.0

~~30.0 1 2.291 1.779 0.392 9.872-2 2 39.662 26.316 7.647 5.968-5 3 386.350 329.620 1.611 0.0~~

~~40.0 1 2.291 1.780 0.392 9.872-2 2 39.750 26.359 7.655 5.965-5 3 392.980 335.340 1.637 0.0~~

~~45.0 1 2.291 1.780. 0.392 9.872-2 2 39.791 26.377 7.658 5.964-5 3 396.420 338.270 1.650 0.0 60.0 1 2.291 1.780 0.392 9.871-2 2 39.949 26.453 7.675 5.962-5 3 407.510 348.170 1.693 0.0~~

~~195 F-1.12~~

~~Tabl . Heave10 y Wate Graphitd ran e Reflector Constants Energy Heavy Water Reflector Graphite Reflector Group 1 1.334 8.801-5 1.409 4.140-6 2 1.247 4.439-6 0.874 1.205-5 3 0.805 8.363-5 0.843 2.430-4~~

~~Tabl . Diffusioe11 n Coefficients From EPRI-CELL Calculation for Core Cross-Section Zert sa o 235U Burnup. Energy 93% Enr. Enr% 20 . Grout D D 1 1.363 1.359 2 1.316 1.295 3 0.893 0.893~~

235 Table 12. EPRI-CELL kœ vs. U Burned for Three Enrichments % 45 % 93 235U Burned ,g 20% 0.0 1.8831 1.8382 1.7827 20.0 1.7429 1.7030 1.6538 30.0 1.7222 1.6829 1.6347 40.0 1.6978 1.6599 1.6128 45.0 1.6837 1.6465 1.6007 60.0 1.6363 1.6055 1.5613

~~Table 13. Values of keff from X-Y Diffusion Theory Calculations Ap, % Enrichment Description keff (BOC-EOC)~~

~~93% Fresh Fuel 1.3162~~

~~93% BOG 1.1800 2.32 93% EOC 1.1485~~

~~45% BOG 1.1674 2.25 45% EOC 1.1375 20% BOG 1.1538 2.00 20% EOC 1.1278~~

~~196 F-1.13~~

~~Fig. 1. 10 MW Heavy Water Benchmark BOC Fluxes at Core Mid-Plane for 93% Enriched Fuel~~

~~8- 6~~

~~_ fas_ t S het i rp m....e a. I _ therma. l~~

~~CO OJ~~

~~3. oo 2S& 50.0 75.0 100.0 125.0 X-AXIS (cm)~~

~~Fig . 2 10-M. W Heavy Water Benchmar k- 45%/93 % BOC Flux Ratios t Cora e Mid-Plane~~

~~•*> s- o l i o04 o xü- u. _^'' ,""" in •>-SOt Xo'~~

~~___ f09t ..... «p{thermal ... thermal~~

~~0.0 25J3 50.0 07& 100.0 125.0 X-AXIS (cm)~~

~~197 F-1.14~~

~~Figure 3. 10 MW Heavy Water Benchmark- 20%/93% BOG Flux Ratios at Core Mid-Plane.~~

~~s- ratO~~

~~CM N~~

~~__ fast s .... «so it her mat 0 __ thermal~~

~~0.0 25.0 50.0 75.0 100.0 125.0 X-AXIS (cm)~~

~~Table 14. Delayed-Neutron-Dependent Parameters~~

~~Enrichment keff Seff A, usée 93* 1.1516 0.007063 460.2 20X 1.1315 0.006890 430.0~~

~~Tabl . 15 eDelaye d Neutron FamilieParametery b 0 d san X s~~

~~932 Enrichment 20Z Enrichment~~

~~1 Families *1, sec" 0i k, sec-1 *i 1 0.0127 2. 7011 x lo-* 0.0127 2. 6295 x 10-* 2 0.0317 1.4960 x 10-3 0.0317 1.4643 x 10-3~~

3 0. 1160 1.3290 X 10-3 0.1163 1.2983 x 10-3 4 0.3110 2.8784 X 10-3 0.3113 2.8009 x 10-3 5 1.3999 9.0529 X 10-* 1.3986 8.8248 x 10-* 6 3. 8692 1.8396 X 10-* 3.8572 1. 8078 x 10-*

~~198 F-1.15~~

~~Tabl . Isothermae16 l Reactivity Feedback Coefficients Enrichmen% 93 t~~

Coolant Fuel D20 H2Û Temp, Temp. , Density, Density, 3 °C °C g/cm g/cm keff* Ap, % Slope 20 20 1.105 0.998 1.15413 - - 50 50 1.096 0.988 1.14897 -0.389 -1. 30 x KrVc 20 300 1.105 0.998 1.15410 -0.007 ~ 2 x 10-8/°C 20 20 0.884** 0.798** 1.15086 -0.246 -1. 23 x io-4/%

~~20% Enrichment 20 20 1.105 0.998 1.13393 50 50 1.096 0.988 1.12860 -0.417 -1. 3x l~~

~~Table 17. Isothermal Reactivity Feedback Components for 20-5 TemperaturC ° 0 e Increas n Coolani e Fued an tl~~

93% Enrichment Coolant Fuel D20 H2Û Temp, Temp. , Density, Density, 3 "C °C g/cm g/cm keff* Ap, % Slope 20 20 1.105 0.998 1.15413 - - 50 20 1.105 0.998 1.15014 -0.301 -1.00 x 10-Vc 20 20 1.096 0.988 1.15284 -0.090 - 7 .32 x 10~4/°C 20 50 1.105 0.998 1.15412 -0.003 ~ 1 x 10-7/-C -0.399 io-*/°x -2 1 .3 c

~~20% Enrichment 20 20 1.105 0.998 1.13393 50 20 1.105 0.998 1.13043 -0.270 - 3 .91 x i~~

~~JUoderator temperature was 25 °C. **Coolant density decreased by 20% in central two fuel elements only.~~

~~199 F-1.16~~

~~Table 18~~

~~Compariso f Absorptioo n n Reaction Rate Ratiod an s s for Specified Isotopes with 93% and 20% Enrichment (based on concentrations of 10~20 at/bn*cm for each isotope)~~

~~Absorption Reaction Rates Care Center~~

Group 235u 98HO 10B Co Si H 1 1.90-6 6.61-8 2.06-6 2.02-8 4.10-9 3.66-8 93Z 2 4.56-5 7.41-7 1.83-4 8.95-6 7.68-9 9.03-8 3 7.82-4 1.51-7 4.55-3 4.44-5 1.91-7 2.2S-6 Total 8.30-4 9.58-7 4.74-3 5.34-5 2.03-7 2.38-6

~~1 1.91-6 6.63-8 2.07-6 2.03-8 4. 1 1-9 3.67-8 20Z 2 4.41-5 7.49-7 1.73-4 8:94-6 7.27-9 8.54-8 3 6.52-4 1.27-7 3.80-3 3.71-5 1.60-7 1.88-6 Total 6.98-4 9.42-7 3.98-3 .4.61-5 1.71-7 2.00-6~~

~~Ratio (20/93) 0.841 0.983 0.840 0.863 0.842 0.840~~

~~Absorption Reaction Rates intm m o 6 Reflecto7 • r~~

~~Group 23Su 98Mo 10B Co SI N~~

~~1 2.59-7 9.35-9 2.97-7 2.85-9 4.29-10 3.36-9 93Z 2 1.05-5 1.66-7 4.35-5 2.03-6 1.82-9 2.14-8 3 7.89-4 1.50-7 4.52-3 4.41-5 1.90-7 2.23-6 Total 8.00-4 3.25-7 4.56-3 4.61-5 1.92-7 2.25-6~~

~~1 2.68-7 9.69-9 3.07-7 2.95-9 4.45-10 3.48-9 20Z 2 1.07-5 1.70-7 4.43-5 2.07-6 1.86-9 2.18-8 3 7.40-4 1.41-7 4.25-3 4.14-5 1.79-7 2.10-6~~

~~Total 7.51-4 3.21-7 4.29-3 4.35-5 1.81-7 2.13-6~~

~~Ratio (20/93) 0.939 0.988 0.941 0.944 0.943 0.947~~

~~The thermal flux «0.62enrichmenZ 93 ) rati d 5eV oan Z betweet 20 case s e th nswa 0.85 at the core center and 0.94 at 76 mm into the reflector.~~

~~200 APPENDI2 XF-~~

~~BENCHMARK SOLUTION~~

~~HARWELL~~

~~An approac Benchmare th o ht k Problem base computinn do g Harwellt a method e us resulte .n Th si calculationf so s performed "by members of the Materials Physics Division, Harwell.~~

~~201 Appendix F2 Benchmark Solution~~

~~Introduction~~

The specification of the Benchmark Problem, which appears in Appendix FO gives adequate scope for proving the ability of the various participants to calculat essentiae th e l reactor physics parameter f theiso r individual heavy water reactors using thei preferren row d computing method d datan s a bases.

~~This paper calculationK giveU resulte e th sth benchmare f th so n so k problem.~~

The reactor physics computing code used most extensively at Harwell is WIMSE the VVinfrith ^Improved Multigroup Sicheme suita , f linkeeo d reactor physics codes which was.evolved for light water, heavy water and graphite moderated thermal reactors varioue Th . s code thin i s s suite intercommunicata a evi serie f interface'so s carrying intermediate results suc neutros ha n spectra, number densities, cross section data, geometry specifications etc., written on disc and read as required in subsequent stages. 'Such functions as lattice cell calculations using collision probability methods, spectrum weighting and cross section averaging, group condensation, editing of spectra to produce reaction rates, collision probability and neutron transport solutions in one dimensional geometry, diffusion calculations in two and three dimensions, burnup calculations etc. can all be performed consecutively by simple commands.

e firsTh benchmare t th tas f o k k exercis perforo t s i e bura mp n-u computation on the infinite lattice cell to observe the variation in kœ during the fuel life. A lattice cell calculation is carried out on the clean fuel in a coarse geometrical mesh in the standard WIMS-E 69 energy group structure. Usin spectrue th g m generate startina s a d g guess thi repeates i s d

Ifinena r k mescalculated d han . Condensatio 4 grou1 a po nt spectru e th n i m 00 fuel regions burn-ue followth r p sfo computatioe phasth f eo whicn i ne th h fission product and actinide inventories are integrated numerically over a numbe verf ro y abort time steps, assumin fissioga n rate equivalen 0 Mw/21 o t 6 per fuel element. At the end of this the code returns with its newly computed

~~202 composition of fuel and fission products to the initial stage, recompute k^~~

~~and neutron spectr nexe a th readt r applicatiofo y burn-ue th f no p modulee Th .~~

~~choice of the length of time-step for the integrations in the burn-up module~~

~~and interval between spectrum recalculation is at the user's command. • e resultfirse Th th tf so tas e show kar Fign i n wher.1 e infinitth e e~~

~~lattice cell multiplicatio plottes ni functioa mase s th f a do s f o n 235.~~

~~% enrichmenburned-u20 d an botr % tpfo h93 fuel e timTh .e dependent concentrations 235 of some of the isotopes of interest are given as a function U burned-up in~~

~~Tables 1 and 2 for 93% and 20% enrichment respectively.~~

~~Table 1~~

~~Concentration selectef so d isotopes during fue l% enrichmen 93 lif t ea t~~

~~U235 Burnt/g Xe 135 Sml49 U235 U236 U238~~

0 0.0 0.0 1.302E-03 0.0 9.679E-05 20 1.283E-08 1.091E-07 1.127E-03 2.848E-05 9.593E-05 30 1.191E-08 1.078E-07 1.041E-03 4.250E-05 9.550E-05 40 1.098E-08 1.055E-07 9.545E-04 5.639E-05 9.507E-05 45 1.051E-08 1.039E-07 9.119E-04 6.319E-05 9.485E-05 60 9.095E-09 9.780E-08 7.829E-04 8.351E-05 9.418E-05

~~Ü235 Burnt/g 9 Pu23 Pu240 PU241 Pu242~~

0 0.0 0.0 0.0 0.0 20 5.878E-07 2.958E-08 .755E-02 9 6.926E-11 30 8.162E-07 6.266E-08 8.657E-09 3.497E-10 40 .003E-01 6 1 .048E-07 1.905E-08 1.106E-09 45 1.081E-06 1 .284E-07 2.599E-08 1.763E-09 60 1 .255E-06 2.071E-07 5.415E-08 5.570E-09

~~203 Table 2~~

~~Concentratio f selecteo n d isotopes during fue lenrichmen1 20 lif t a e t~~

~~U235 Burnt/g Xe 135 Sal 4 9 U235 U236 U238~~

0 0.0 0.0 1.450E-03 0.0 5.710E-03 20 1.470E-08 1.250E-07 1.278E-03 2.874E-05 5.694E-03 30 1.390E-08 1.252E-07 1.192E-03 4.292E-05 5.686E-03 40 1.308E-08 1.244E-07 1.106E-03 5.699E-05 5.679E-03 45 1.266E-08 1.236E-07 1 .063E-03 6.395E-05 5.675E-03 60 1.137E-08 1.198E-07 9.341E-04 8.465E-05 5.662E-03

~~U235 Burnt/g Pu239 Pu240 PU241 Pu242~~

0 0.0 0.0 0.0 0.0 20 1.224E-05 5.576E-07 4.946E-08 1.105E-09 30 1.731E-05 1.194E-06 1.573E-07 5.600E-09 40 2.171E-05 2.019E-06 3.498E-07 1.772E-08 45 2.366E-05 2.491E-06 4.815E-07 2.838E-08 60 2.852E-05 4.108E-06 1.025E-06 8.994E-08

The next task in the benchmark calculations involves a whole reactor model with typica lcyclf o beginnin d e en (BOd E0Can d gC an ) distributiond an w ne f so partially burned-up fuel element detailes a s WIMS e th Appendin En i dI . FO x suite this was tackled by generating the diffusion coefficients for the individual fuel element averaginy sb reactioe th g n rates ovee infinitth r e lattice cell usin WSMEAe th g H module. Module COL thes i L n employe carro t d y out a group condensation from the original 69 to the more manageable 7 groups for performin finae gth l stagegeometrY X e ,th y calculatio reactoe th f no r

~~witSNAe hth P module resulte k-effective Th .th f so e calculationC BO o f s conditionC andEO givee sar Tabln i n . e3~~

~~204 Table 3~~

~~> , * ~ ..., ^ , ' « reactivity loss Enrichment Condition k-effective . Ap (BQC _EQC )~~

~~93% BOC 1.143 2.68% 93% EOC 1.109~~

~~20% BOC 1.127 2.12% 20% EOC 1.101~~

~~Since an important consideration in enrichment reduction of the fuel supplie deleteriouse th lie n si s effect neutron so n flux levelse , th par f o t~~

~~benchmark problem was a comparison of these at 93% and 20% enrichment.~~

~~Fig. 2 shows the flux distribution along the X-axis of the XY geometry~~

~~reactor computatio energ3 n ni y group 0.62- 0.62, 5.5s0 - SeV V V Se 3ke~~

~~and 5.53 keV to 10 MeV for the 93% enrichment case, while Fig. 3 shows the~~

~~ratio of the fluxes in each of the 3 groups between the 20% and 93% enriched~~

~~reactors. It can be seen that enrichment reduction from 93% to 20% is most~~

~~keenl thermae y th fala fel n s i la t lcoree mora th flu r less eo ,n a i x s~~

~~direct consequence of the increase in fuel content of the fresh elements from~~

~~150 g to 167.5 g 235U.~~

~~The final part of the benchmark problem addresses the reactor safety area~~

~~anconcernes i d d with reactor temperatur void an ed coefficient% 93 e th t a s~~

~~and 20% enrichment levels with the original 150 and 167.5 g 235 U respective~~

~~fuel element loadings subjecte burn-uU unifora o g t dp 0 m4 acros coree sth . 235~~

~~The perturbations to the base 20 C uniform temperature case were:~~

~~o o 1 uniform temperature change from 20 C to 50 C for all fuel and coolant, 2 uniform temperature change from 20 C to 300 C for the fuel plates but~~

~~wit expansioo hn n effect, 3 20% reduction in coolant density for the two central fuel elements.~~

~~205 The coolant was taken to be all the heavy water inside the outer aluminium~~

~~tube. That betwee fuee th n l reflectoe elementth d core d an th an e O n rD i s~~

graphit densitiee 20°t 50°d a Th C an eCO D . remainewer C f so 0 e 2 taket a d n -3 -3 to be 1.104g cm and 1.094g cm respectively. * e resultTh s obtaine reactivitr fo d givee e colum ar yTh Tabln i n n. e4

~~labelle ^ give k dvalue th s e obtained frolattice th m e calculation value Th . e~~

~~for k . and the change in reac-tivity, Ap, are from X-Y geometry diffusion~~

~~calculations. Reactio principae n th rate r fo s l nuclide energ3 n si y groups~~

~~(boundaries at 5.53 keV and 0.625 eV) are given in Tables 5 to 8 for the~~

~~4 cases. The reaction rates have been obtained from lattice calculations~~

œ spectrumk wit a h . Tabl givee9 diffusioe sth n coefficient componente th n si s of the XY calculations, including the various 40g burned up smeared core cells and heavy water and graphite reflectors.

~~Table 4~~

~~enrichmen% 93 t 20% enrichment Cssc k» keff AP* ~ k * Ap keff~~

Base, all at 1.696169 1.129467 0 1.620118 1.115849 0 20°C 50 C coolant moderator and 1.694302 1.123995 -0.431 1.617244 1.092113 -0.424 fuel 300°C fuel 1.696003 1.129242 -0.018 1.610962 1.110259 -0.451 20% coolant void 1.694458 1.12598 -0.274 1.617476 1.112561 -0.256

~~206 TABLE 5~~

~~Reaction Rates For Standard Case~~

~~93% case 20% caae Nuclide Group v-fission fission capture \» -fission fission capture~~

235 u 1 0.004702 0.001875 0.002316 0.005447 0.002172 0.002683 2 0.070826 0.029096 0.044543 0.077346 0.031774 0.048598 3 1.61715 0.66S391 0.782152 1.46416 0.602436 0.708319 1-3 1.69268 0.696400 0.82901 1.54695 0.63638 0.759600

238 u 1 0.000031 0.000011 0.000029 0.002225 0.000791 0.002080 2 0.000000 0.000000 0.002463 0.000000 0.000000 0.047322 3 •0.000000 0.000000 0.000275 0.000000 0.000000 0.015335 1-3 0.000031 0.000011 0.002768 0.002225 0.000791 0.064737

239 1 0.000006 0.000002 0.000002 0.000135 0.000045 0.000054 Pu 2 0.000086 0.000030 0.000052 0.001975 0.000688 0.001184 3 0.003504 0.001221 0.001779 0.068237 0.023772 0.034804 1-3 0.003596 0.001253 0.001833 0.070347 0.024505 0.036042

~~Al .1 0.000991 0.000868 2 0.004472 0.003772 3 0.068109 0.047676 1-3 0.07357 0.052317~~

~~D 1 -0.003691 -0.003705 2. 0.000074 0.000071 3 0.002295 0.001856 1-3 -0.001322 -0.001778~~

~~R 1 0.000000 0.000000 2 0.000122 0.000116 3 0.003783 0.003059 1-3 0.003905 0.0031755~~

~~0 1 0.003536 0.003550 2 0.000002 0.000001 3 0.000200 0.000162 1-3 0.003738 0.003713~~

~~207 TABLE~~

~~Reactio CoolanC n 0 Rate5 r t fo sModerato Fued an rl~~

~~93% case 20% case~~

~~Nuclide Group y-fission fission capture V-fission fission capture~~

»"u 1 0.004119 0.001636 0.002019 0.004772 0.0018948 0.002338 2 0.072658 0.029848 0.045811 0.061505 0.025266 0.037412 3 1.61204 0.663282 77993. P 6 1.27855 0.526061 0.618728 1-3 1.68882 0.694766 0.827765 1.34483 0.553221 0.658478

238 u 1 0.000031 0.000011 0.000035 0.002235 0.000795 0.002446 2 0.000000 0.000000 0.002599 0.000000 0.000000 0.170324 3 0.000000 0.000000 0.000276 0.000000 0.000000 0.013472 1-3 0.000031 0.000011 JO. 00291 0.0022345 0.000795 0.186242

239Pu 1 0.000006 0.000002 0.000002 0.000136 0.000045 0.000054 2 0.000087 0.000030 0.000052 0.001786 O. 000622 0.001072 3 0.003595 0.001252 0.001835 0.061238 0.021333 0.031401 1-3 0.003688 0.001285 0.00189 0.06315 0.022001 0.032528

~~Al 1 0.000997 0.000873 2 0.004509 0.003471 3 0.06815 0.041766 1-3 0.073656 0.046110~~

~~D 1 -0.003689 -0.003703 2 0.000074 0.000063 3 0.002259 0.001599 1-3 -0.001356 -0.00204~~

~~H 1 0.000000 0.000000 2 0.000122 0.000104 3 0.003724 0.002637 1-3 0.003845 0.002740~~

~~0 1 0.003535 0.003548 2 0.000001 0.000001 3 0.000197 0.000139 1-3 0.003733 0.003688~~

~~208 TABLE?~~

~~Reaction Rates for 300 C Fuel~~

~~93% case 20% case~~

~~Nuclide Group v-fission fission capture v-fission fission capture~~

235 u 1 0.004095 0.001626 0.002006 0.004744 0.001884 0.002324 2 0.070846 0.029104 0.044555 0.076907 0.031593 0.048317 3 1.61757 0.665565 0.782357 1.45591 0.599044 0.704331 1-3 1.69251 0.696295 0.828919 1.53757 0.632521 0.754971

238 0 1 .0.000031 0.000011 0.000034 0.002225 0.000791 0.002442 2 . 0.000000 0.000000 0.002506 0.000000 0.000000 0.052653 3 0.000000 0.000000 0.000275 0 . 000000 0.000000 0.015249 1-3 0.000031 0.000011 0.002816 0.002225 0.000791 0.070343

239Pu 1 0.000006 0.000002 0.000002 0.000135 0.000045 0.000054 2 0.000086 0.000030 0.000052 0.001965 0.000685 0.001178 3 0.003505 0.001221 0.001779 0.067851 0.023637 0.034607 1-3 0.003597 0.001253 0.001833 0.069951 0.024367 0.035839

~~Al 1 0.000991 0.000868 2 0.004473 0.003756 3 0.068127 0.047408 1-3 0.073592 0.052032~~

~~D 1 -0.003691 -0.003704 2 0.000074 0.000070 3 0.002295 O.O01845 1-3 -0.001322 -0.001789~~

~~B 1 0.000000 0.000000 2 0.000122 0.000115 3 0.003784 0.003042 1-3 0.003906 0.003158~~

~~0 1 0.003536 0.003549 2 0.000002 0.000001 3 0.0002 0.00016 1-3 0.003738 0.003712~~

~~209 TABLE 6~~

~~Reaction Rajfces for 20% Void in Coolant~~

~~93% case cas% 20 e~~

~~Nuclide Group v-fissîon fission capture \>-fission fission capture~~

235 u 1 0.004974 0.0019S3 0.00245 0.005760 0.002297 0.002837 2 0.074932 0.030782 0.04711 0.081786 0.033566 0.0513210 3 1.61104 0.662878 0.779269 1.45558 0.598904 0.704238 1-3 1.69095 0.695643 0.828829 1.54305 0.634767 0.758394

238 u 1 0.000033 0.000012 0.000031 0.00235 0.000839 0.002201 2 0.000000 0.000000 0.002597 0.000000 0.000000 0.048142 3 0.000000 0.000000 0.000275 0.000000 0.000000 0.015265 1-3 0.000033 0.000012 0.002903 0.00235 0.000839 0.065608

239 Pu 1 0.000006 0.000002 0.000002 0.000143 0.000048 0.000057 2 0.000091 0.000032 0.000055 0.002091 0.000728 0.001253 3 0.003528 0.00123 0.001794 0.068616 0.023904 0.035068 1-3 0.003626 0.001263 0.001851 0.07085 0.024679 0.036378

~~Al 1 0.001052 0.000922 2 0.004740 0.003994 3 0.067768 0.047314 1-3 0.07356 0.05223~~

~~D 1 -0.003668 -0.003682 2 0.000074 0.000070 3 0.0216716 0.001751 1-3 -0.001427 -0.001861~~

~~H 1 0. 000000 0.000000 2 0.000121 0.000115 3 0.003573 0.002887 1-3 0.003694 0.003002~~

~~0 1 0.003514 0.003528 2 0.000002 0.000001 3 0.000189 0.000153 1-3 0.003704 0.003682~~

~~210 TABLE~~

~~Diffusion Coefficients~~

~~Enrichmen% 93 t Enrichmen% 20 t Case Dl D2 D3 Dl D2 D3~~

~~t Basea l ,al 20°C 1.51734 1.33954 0.92203 1.52367 1.32232 0.92244 0 coolant5 , moderator 1.52302 1.34407 0.93419 1.52738 1.32514 0.93172 and fuel~~

~~300°C fuel• 1.51739 1.33954 0.92203 1.52367 1.32178 0.92244 coolan% 20 t void 1.63309 1.42160 0.97480 Heavy water reflector 1.47086 1.26695 0.897424 Graphite reflector 1.40215 0.87336 0.84280~~

~~211 Fig 1 .Variatio n Infiniti n e Lattice Cell multiplication witU burn-uh p 235~~

~~1.90~~

~~1.80~~

~~1.70~~

~~1.60~~

~~1.50 10 20 30 40 50 60~~

~~212 Pig2 . Hadtal Distributio Grou3 f o np Fluxe Cort a s e Mldplan% enriche 93 w reactoM C r 0 BO fo e 1 dt a r~~

~~120 Radial Distance from Core Centre. CD Pig. 3 Radial Distribution of Flux Ratios 0(20)/0(93)~~

~~1.05~~

~~1.0~~

~~-9S~~

~~40 90 0 13 100 Ola from Core Centre,~~

~~214 APPENDIX F-3~~

~~AUSTRALIAN BENCHMARK CALCULATIONS~~

~~G.S. ROBINSON~~

~~Nuclear Technology Division~~

~~Australian Atomic Energy Commission~~

~~Research Establishment~~

~~Lucas Heights~~

~~215 INTRODUCTION~~

Calculations of the methodical benchmark problem have been undertaken using methods developed at the AAEC and incorporated in the AUS modular code scheme (Robinson 1975a). These calculations for HEU and LEU fuel consist of cell DIDO-type burnuth r pfo e fuel element diffusiod ,an n theory applien a o dt X-Y representation of reactor core. Brief descriptions of the computer codes usetheid an d r applicatio thio nt s proble givene mar , together with results fueU LE l d elemenan foU rHE t cell burn-u 2-dimensionad pan l diffusion theory (X-Y) reactor core calculations.

~~METHOD OF CALCULATION AND RESULTS~~

Generation of Cross Sections for Core and Reflector Cross sectionfunctioa core s th ea r burn-uf sfo no p were generated using the four AUS modules MIRANDA, ANAUSN, EDIT and CHAR, which are described briefl turnn yi radiaA . l fuee modeth l f lelemeno t wit circularisedha , reflective (white) outer boundar useds ywa .

MIRANDA (Robinson 1977) is a data preparation code which generates multigroup cross sections for each of the materials in a lattice cell. The cod uses ewa d wit 128-grouha p cross section library derived from ENDFB-IV which includes resonanc efore subgrouf datth mo n ai p parameters. The resonance calculatio subgroua s ni p method which uses collision probability routines to represent spatial effects. For these problems, nine regions were resonance th use n i d e calculation, wit foue hth r fuel-meat tubes represented explicitly. This is considerably more complex than the more usual 3-region (fuel, can, coolant) resonance calculation used in MIRANDA. Some energy condensatio alss nwa o performe MIRANDAn di , which include homogeneousa s spectrum calculation, to produce a 25-group cross section set, of which 15 groups were belo. eV w1

The ANAUSN module (B.E. Clancy, AAEC unpublished) is a general purpose, one-dimensional discrete ordinate program with anisotropic scattering whics hi simila ANISo rt N (Engle uses 1967)wa d isotropit n I hera . n ei c scattering, S4, 29 mesh interval calculation (maximum interval of 6 mm) using the 25-group date from MIRANDA. All materials were represented explicitly.

The EDIT module (J.P. Pollard, AAEC unpublished uses )edio wa de t tth flux outpu ANAUSf to foro Nt m cell average cross sections perforo ,t ma homogeneous flux calculation which included buckling to give k = 1.08, and to condense to 5-group cross sections over the near critical spectrum. The energy boundaries used were 0.821 MeV, 5.531 keV and 0.618 eV.

216 The CHAR module (Robinson 1975b) is a raultiregion burn-up module which uses analytic technique perforo st m nuclide burn-up using spatial fluxes from other AUS modules. In these calculations, the ANAUSN fluxes adjusted to the k = 1.08 calculation were used to form the nuclide depletion equations for 6££ foue th eacr f fuel-meaho t tubes cele Th .l power leve average lth use s edwa value for the core. All results are given with a Xenon level corresponding to this average core power.

Repeated cycles through the above four modules were made to generate 5-group cross section functioa s sa burn-upf no . Cross-section heave th yr sfo wate graphitd ran e reflectors were generated separatel MIRANDn yi y Ab 235 condensation over the flux obtained from a U fission sourc eacn ei h material. fuelU eacr LE sfo d h an burn-u U œ valueK HE r p sfo poin whict ta h lattice calculations were performed are given in Table 1. Reaction rates for some important nuclides in the K^ spectrum at zero burn-up are given in Table 2. 135 149 Number densities for uranium and plutonium isotopes, Xe and Sm as a U fueLE l d functioan U HE burn-uf no r fo givee 4 par Table n d ni an s3 respectively. The values given have been obtained by linear interpolation between burn-up points actually calculated.

Global Calculations XY flux calculations were performed in the POW module (Pollard 1974) whictwo-dimensionala s hi , finite difference diffusion code using edge mesh points. Linear interpolatio mas U uses givso wa d t n eni cros23s5 sectiont sa the required burn-up values 5-groupe Th . , quarter-core diffusion calculations were performed wit interval2 mes3 ha f interval6 ho 1 19.0d f so an f m 5so m 38.1 mm in the X direction. A similar mesh was used in the Y direction. A uniform axial bucklin includeds 8.71f wa go 2 valuee 8m~ Th . s obtainer dfo

~~keff~~

~~eff Fuel Enrichment BOC EOC~~

~~HEU% ,93 1.1696 1.1368~~

~~LEU% 20 , 1.1449 1.1194 1~~

The fluxe core s th axialonX e t e midplansa g th e (assumin choppega d cosine axial distribution thren )i e energC BO corU HE t ye a e groupth r sfo are give groue Figurn Th ni p . boundariee1 thir sfo s edi 5.53e tar V 1ke and 0.618 eV. The ratio of fluxes in the LEU core at BOC to those in the

217 givee HEquotear e Figurn G UnTh i BO cor d. et 2 ea fluxe s have been obtained by normalisin resulte non-criticae gth th f so l calculations directle th o yt required power level. The difference in k _,. is thus responsible for most of the differences in the fast and epithermal flux in Figure 2, In the LEU core, the thermal flux peaks on the X axis are at the centre (1.27 x 101 4 ) and at 522 mm (1.24 x 1014). Reactivity Coefficient Calculations

These calculations have been performe identican a n di l mannee th o rt burn-up calculations reactivite Th . y changes have been obtained from direct XY diffusion calculations using a tight convergence criterion (10 ). The base case was the same as the previous global calculations but with 235 aelementr uniforpe threU e Th .me g burn-uperturbation 0 4 f po s considered were: 1. uniform temperature change from 20°C to 50°C for all fuel and coolant, 2. uniform temperature change from 20°C to 300°C for the fuel plates but with no expansion effect, reductio% 20 3. coolann ni centrao ttw densite lth r fueyfo l elements. The coolant was taken to be all heavy water inside the outer aluminium tube.

~~The results obtaine reactivitr dfo givee valuey ^ ar Tablk n ni e . e5 sar obtained from the lattice calculation, k values and the changes in reactivity, e froAp diffusio,ar Y m X n calculations.~~

Reaction rates for the principal nuclides in three energy groups (boundaries at 5.53 keV and 0.618 eV) are given in Tables 6 to 9 for the four cases. The reaction rates have been obtained from lattice calculations with a k spectrum and have been normalised to unit neutron loss (i.e. fission + capture - (n,2n) = 1).

~~Diffusion coefficient givee havsd ar Tabl n an ni e 0 alse1 o been obtained from the k spectrum calculations. CO~~

~~218 REFERENCES~~

~~Engle, W.W. (1967 user'A )- s manua ANISr one-dimensionaa lfo N- l discrete ordinates transport code with anisotropic scattering. K-1693.~~

~~Pollard, J.P. generaa (1974 modulS - AU W D )- 2 lePO d purposan 1 e0, multigroup neutron diffusion code including feedback-free kinetics. AAEC/E269.~~

~~Robinson, G.S. (1975a) - AUS - the Australian modular scheme for reactor neutronic computations. AAEC/E369.~~

~~Robinson, G.S. (1975b) - AUS burn-up module CHAR and the associated STATUS data pool. AAEC/E372.~~

~~Robinson, G.S. (1977) - AUS module MIRANDA - a data preparation code based on multiregion resonance theory. AAEC/E410.~~

~~TABLE 1~~

~~k AS A FUNCTION OF BURN-UP~~

~~Enrichmen% 93 t 1 20% Enrichment i grams 235U k grams 235U k burnt OO burnt CO l 0 1.8899 1 n 1.7905~~

~~0.50 1.8015 0.50 1.7085~~

~~10.56 1.7670 10.60 1.6767~~

~~20.60 1.7465 20.57 1.6580~~

~~30.58 1.7243 3C.41 1.6381~~

~~40.51 1.6995 40.13 1.6165~~

~~50.39 1.6718 49.73 1.5933~~

~~60.20 1.6402 59.19 1.5683~~

~~69.94 1.6037 68.51 1.5409~~

~~219 TABLE 2~~

~~REACTION RATE ZERT SA O BURN-UN I P LATTICE CALCULATIONS~~

* * Fractional Events for 93% Case Fractional Events for 20% Case Nuclide Group v-fission fission capture v-fission fission capture

1 0.00539 0.00215 0.00056 0.00600 0.00240 0.00062 2 0.09477 0.03918 0.01980 0.10054 0.04156 0.02062 U 3 1.78977 0.73994 0.12724 1.68206 0.69541 0.11978 1-3 1.88992 0.78127 0.14759 1.78860 0.73937 0.14102

~~1 0.00003 0.00001 0.00002 0.00195 0.00069 0.00143 2 0.0 0.0 0.00285 0.0 0.0 0.04794 238u t 3 0.0 0.0 0.00027 0.0 0.0 0.01343 1-3 0.00003 0.00001 0.00314 0.00195 0.00069 0.06280~~

~~1 0.00107 0.00103 2 0.00291 0.00265 Al 3 0.05902 0.04827 1-3 0.06299 0.05195~~

~~1 0.00299 0.00298 2 0.00009 0.00008 D2O T 3 0.00225 0.00194 1-3 0.00533 0.00500~~

~~1 0.00001 0.00001 2 0.00013 0.00013 H20 3 0.00342 0.00295 1-3 0.00356 0.00309~~

* Normalized to unit loss. That is fissio capturn+ e(n,2n- 1 = )

~~t also (n,2n 0.0000f )cas% o 20 e n 4i~~

~~t also (n,2n) of 0.00388 and 0.00387 in 93% and 20% cases respectively~~

~~220 TABLE 3~~

~~NUMBER DENSITIES FOR HEU FUEL (Nuclei per 10~24 cm3)~~

~~U235 Burnt (g) Xe 135 Sml49 U235 U236 U238~~

0 0.0 0.0 1.302E-03 0.0 9.679E-05 20 1.291E-08 9.839E-08 1.129E-03 2.809E-05 9.600E-05 30 1.197E-08 9.580E-08 1.042E-03 4.190E-05 9.559E-05 40 1.102E-08 9.303E-08 9.551E-04 5.555E-05 9.517E-05 45 1.054E-08 9.140E-08 9.117E-04 6.229E-05 9.496E-05 60 9.103E-09 8.588E-08 7.814E-04 8.233E-05 9.431E-05

~~U235 Burnt (g) PU239 PU240 PU241 PU242~~

0 0.0 0.0 0.0 0.0 20 6.710E-07 3.493E-08 3.627E-09 8.566E-11 30 9.335E-07 7.321E-08 1.117E-08 4.204E-10 40 1.149E-06 1.216E-07 2.430E-08 1.309E-09 45 1.236E-06 1.493E-07 3.373E-08 2.215E-09 60 1.450E-06 2.384E-07 6.856E-08 6.553E-09

~~221 TABLE 4~~

~~NUMBER DENSITIE 3 FUEU cm LE LR (NucleSFO 0 1 r ipe -24~~

~~U235 Burnt (g) Xel35 Sml49 U235 U236 U238~~

0 0.0 0.0 1.454E-03 0.0 5.744E-03 20 1.485E-08 1.131E-07 1.281E-03 2.831E-05 5.728E-03 30 1.404E-08 1.118E-07 1.194E-03 4.227E-05 5.719E-03 40 1.321E-08 1.104E-07 1.107E-03 5.608E-05 5.711E-03 45 1.279E-08 1.094E-07 1.064E-03 6.290E-05 5.706E-03 60 1.149E-08 1.057E-07 9.334E-03 8.321E-05 5.692E-03

~~U235 Burn) (g t PU239 PU240 PU241 PU242~~

0 0.0 0.0 0.0 0.0 20 1.369E-05 6.516E-07 6.455E-08 1.357E-09 30 1.935E-05 1.376E-06 2.001E-07 6.626E-09 40 2.423E-05 2.308E-06 4.385E-07 2.054E-08 45 2.631E-05 2.855E-06 6.142E-07 3.491E-08 60 3.175E-05 4.644E-06 1.270E-06 1.042E-07

~~222 TABLE 5~~

~~REACTIVITY COEFFICIENTS~~

~~Case Description Fuel k k ,_ Ap% Ap/°C Type CO eff~~

~~Base HEU 1.70057 1.14962 LEU 1.61652 1.12903~~

~~50 °C coolant and fuel HEU 1.69870 1.14433 -0.403 -1.34 x 10~* LEU 1.61365 1.12398 -0.398 -1.33 x 10~~

~~300°C fuel HEU 1.70039 1.1495 610~x g -0.002 - 5 LEU 1.60730 1.12391 -0.403 -1.44 x 10~~

~~20% coolant void HEU 1.69909 1.14612 -0.266 (central) LEU 1.61370 1.12579 -0.255~~

~~223 TABLE 6 REACTION RATES FOR BASE CASE~~

~~Nuclide Group Fractional Events for 93% case Fractional Events for 20% case v-fission fission capture v-fission fission capture~~

1 0.00396 0.00158 0.00041 0.00457 0.00183 0.00047 235n 2 0.07068 0.02922 0.01484 0.07772 0.03213 0.01604 3 1.62169 0.67045 0.11500 1.46004 0.60362 0.10375 1-3 1.69633 0.70125 0.13025 1.54234 0.63758 0.12027

~~1 0.00003 0.00001 0.00002 0.00195 0.00069 0.00142 238u t 2 0.0 0.0 0.00284 0.0 0.0 0.04826 3 0.0 0.0 0.00032 0.0 0.0 0.01518 1-3 0.00003 0.00001 0.00319 0.00195 0.00069 0.06486~~

~~1 0.0 0. 0. 0.00012 0.00004 0.00001~~

~~239pu 2 0.00010 0.00004 0.00003 0.00213 0.00074 0.00050 3 0.00403 0.00140 0.00065 0.06842 0.02381 0.01124 1-3 0.00414 0.00144 0.00068 0.07067 0.02459 0.01175~~

~~1 0.00107 0.00103 Al 2 0.00292 0.00266 3 0.07213 0.05458 1-3 0.07612 0.05827~~

~~1 0.00299 0.00300~~

~~D2O 2 0.00009 0.00008 3 0.00267 0.00215 1-3 0.00575 0.00523~~

~~1 0.00001 0.00001~~

~~H2O 2 0.00014 0.00013 3 0.00406 0.00327 1-3 0.00421 0.00341~~

~~t (n,2n 0.0000f )% o 20 n 4i case ± (n,2n) of 0.0038% 8 20 an d d an 0 .% 003993 n 0i cases respectively~~

~~224 TABLE 7 REACTION RATES FOR 50°C COOLANT AND FUEL~~

~~Nuclide Group Fractional cas % Event93 e r sfo Fractional Events for 20% case v-fission fission capture v-fission fission capture~~

1 0.00396 0.00158 0.00041 0.00459 0.00183 0.00047 235D 2 0.07091 0.02932 0.01489 0.07791 0.03221 0.01608 3 1.61952 0.66955 0.11498 1.45631 0.60208 0.10361 1-3 1.69439 0.70045 0.13028 1.53881 0.63612 0.12016

~~1 0.00003 0.00001 0.00002 0.00195 0.00069 0.00143 238D t 2 0.0 0.0 0.00285 0.0 0.0 0.04897 3 0.0 0.0 0.00032 0.0 0.0 0.01518 1-3 0.00003 0.00001 0.00320 0.00195 0.00069 0.06558~~

~~1 0.0 0.0 0.0 0.00012 0.00004 0.00001~~

~~239pu 2 0.00011 0.00004 0.00003 0.00214 0.00074 0.00051 3 0.00407 0.00142 0.00066 0.06906 0.02403 0.01142 1-3 0.00418 0.00146 0.00069 0.07132 0.02482 0.01194~~

~~1 0.00107 0.00103 Al 2 0.00293 0.00266 3 0.07231 0.05464 1-3 0.07631 0.05834~~

~~1 0.00299 0.00300 j. D20 2 0.00009 0.00008 3 0.00269 0.00216 1-3 0.00577 0.00524~~

~~1 0.00001 0.00001 H2O 2 0.00014 0.00013 3 0,00408 0.00328 1-3 0.00423 0.00342~~

~~t (n,2n 0.0000 f )o cas% 20 e n 4i * (n,2n) of 0.00388 and 0.00390 in 93% and 20% cases respectively.~~

~~225 TABLE 8 REACTION RATE 3ÛO°R SFO C FUEL~~

~~Nuclide Group Fractional Events for 93% case Fractiona lcas% Event20 e r sfo v-fission fission capture v-fission fission capture~~

1 0.00396 0.00158 0.00041 0.00458 0.00183 0.00047 235u 2 0.07089 0.02931 0.01493 0.07760 0.03208 0.01603 3 1.62131 0.67029 0.11497 1.45136 0.60003 0.10314 1-3 1.69615 0.70118 0.13031 1.53354 0.63394 0.11964

~~1 0.00003 0.00001 0.00002 0.00195 0.00069 0.00143 238u t 2 0.0 0.0 0.00289 0.0 0.0 0.05364 3 0.0 0.0 0.00032 0.0 0.0 0.01509 1-3 0.00003 0.00001 0.00324 0.00195 0.00069 0.07015~~

~~1 0.0 0.0 0.0 0.00012 0.00004 0.00001~~

~~239pu 2 0.00011 0.00004 0.00003 0.00212 0.00074 0.00051 3 0.00403 0.00140 0.00065 0.06802 0.02367 0.01117 1-3 0.00414 0.00144 0.00068 0.07026 0.02445 0.01169~~

~~1 0.00107 0.00103 Al 2 0.00292 0.00264 3 0.07211 0.05426 1-3 0.07610 0.05793~~

~~1 0.00299 0.00300~~

~~D20 2 0.00009 0.00008 3 0.00267 0.00214 1-3 0.00575 0.00522~~

~~1 0.00001 0.00001~~

~~H20 2 0.00014 0.00013 3 0.00406 0,00325 1-3 0.00421 0.00339~~

~~t (n,2n0 f )o ,% 0000cas20 en 4i ± (n,2n) ocase% f20 0sd ,0038 an respectively0.0039 d % 8an 93 n 0i .~~

~~226 TABLE 9 REACTIO VOIN% COOLANN 20 RATEDI R SFO T~~

~~Nuclide Group Fractional cas % Event93 e r sfo Fractiona lcas% Event20 e r fo s v-fission fission capture v-fission fission capture~~

1 0.00420 0.00168 0.00043 0.00486 0.00194 0.00050 235u 2 0.07480 0.03092 0.01569 0.08211 0.03394 0.01693 3 1.61570 0.66798 0.11466 1.45173 0.60019 0.10324 1-3 1.69470 0.70058 0.13079 1.53869 0.63607 0.12067

~~1 0.00003 0.00001 0.00003 0.00208 0.00074 0.00151 238Ö t 2 0.0 0.0 0.00299 0.0 0.0 0.04898 3 0.0 0.0 0.00032 0.0 0.0 0.01511 1-3 0.00003 0.00001 0.00334 0.00208 0.00074 0.06560~~

~~1 0.0 0.0 0.0 0.00013 0.00004 0.00001~~

~~239FU 2 0.00011 0.00009 0.00003 0.00226 0.00079 0.00054 3 0.00406 0.00141 0.00066 0.06889 0.02398 0.01139 1-3 0.00418 0.00145 0.00069 0.07127 0.02480 0.01194~~

~~1 0.00113 0.00109 Al 2 0.00310 0.00281 3 0.07181 0.05422 1-3 0.07604 0.05812~~

~~1 0.00297 0.00298~~

~~D2O 2 0.00009 0.00008 3 0.00253 0.00203 1-3 0.00559 0.00510~~

~~1 0.00001 0.00001~~

~~H2O 2 0.00013 0.00013 3 0.00385 0.00309 1-3 0.00399 0.00323 t (n,2n) of 0.00005 in 20% case ± (n,2n.case% 20 0.0038 f )so d an respectively0.0038 d % 6an 93 n 7i .~~

~~227 TABL0 1 E DIFFUSION COEFFICIENTS~~

~~Case descriptio ncase% 93 sr fo D D for 20% cases~~

~~Group 1 Group 2 Group3 Group1 Group 2 Group 3~~

~~Base 1.4810 1.3179 0.8833 1.4761 1.3040 0.8843~~

~~50°C coolant and 1.4856 1.3213 0.8878 1.4807 1.3072 0.8885 fuel~~

~~300°C fuel 1.4810 1.3179 0.8833 1.4761 1.3031 0.8843~~

~~20% coolant void 1.5933 1.3988 0.9347 1.5877 1.3830 0.9348~~

~~Heavy water refl. 1.4364 1.2464 0.7919~~

~~Graphite refl. 1.3409 0.87473 0.82503~~

228 Flg . 1 Fluxe. t Cora s e tlld-plane 1 EnricheZ 93 i r fo di Fuei l • 1.0 i ' i 1 ' 1 ^ "—-x. Core Graphe t l * D2°* 7 » ^ N CO '*"^ »^ 1.2 v - V x, ' V • ^^^ * £ \ o ^**"***^_ \ M c » ^ \ . N , \ - 0.8 \ o ^•-«»^^ \ •—— — FB8T m s *-< ^^^ *>><^ \ \ —— Epi-ncmn. N^> V ———— TKWflL ^" m ( ^ t ^ • \ X v N, O.i» \ \ — _ . u_ xv m - v\ - x • 0.0 >lil ^•N^"""» J ^ l 0.0 itO.O 80.0 120.0 s Coml ax 3 - X

~~Fig. 2. Flux Ratios at Core Mid-plane 201 / 931 Enriched Fue r 1.1 • 0 i • , | , | , 1 ' 1 •~~

Core D20 Graphe t l 1.05 - 0 "-*- — " • .-'"'*" 4- - O 1.00 ÛC • — X - — • — " _f 0.95 — f - u_ f' . —————— EPI-TÄWBFBST . /' - 0.90 ^y ———— THBm. . _ ^— ,-"* i • 0.85 r f 1,1, i,i, 0.^ 0 0.0 80.0 120.0 X - axls Com3

~~229 APPENDIk XF-~~

~~Benchmark Calculationr fo s~~

~~Heavy Water Reactors~~

~~K. Tsuchihashi~~

~~Japan Atomic Energy Research Institute~~

~~Tokai rbaraki Japan~~

~~231 CALCULATION BENCHMARA R SFO K HEAVY WATER REACTOR USING DIDO-TYPE FUEL ELEMENTS WITH HIGH AND LOW ENRICHMENTS~~

INTRODUCTION A benchmark proble heava r yfo m water reacto s calculaterwa n i d orde o comparrt e calculational methods use varioun i d s research institutes for neutronic parameters. Detailed specifications of the benchmark proble showe ar mAppendin i n x F-0. They, corresponMW 0 1 a o t d 26 element core at beginning (BOO and end (EOC) of the equilibrium cycle d anothe,an r cas uniformlf o e 0 gray4 m burn addes i t o simplift d y the problem. The core is surrounded by an inner reflector of heavy water anouten a d r reflecto graphitef ro . Uraniu% m20 enrichmentd an % 93 f o s are considere DIDO-typn i d e cylindrical fuel elements containing fresh U-235 loading of 150.0 and 167.5 grams, respectively. To perform the neutronic calculations for research reactors, JAERI staff are offered the choice of two computer codes; one the RETER-ACE whic bees ha hn use IAEn i d A benchmark calculation r neutronicsfo d an s safety related parameter MTR-typa r fo s e research reactor othee ,th r bein SRAe th gC (JAERI-M 9781) whic bees ha hn newly developee th r fo d neutroni cJAERe th par If o t thermal reactor standard code systemr Fo . the present benchmark proble e usemSRAe w th o treadCt exac e th t t geometry of fuel element of concentric fuel plates in resonance and thermal calculations by the collision probability method without assuming a unit cell of single fuel plate.

~~METHOD OF CALCULATION AND RESULTS Cross section library~~

e necessarTh y datextractee ar a d fro e SRAmth C fundamental library energe whicth s y hha group structur 7 group10 4 groupf o e(7 s r fas fo st energy range and 45 groups for thermal range, with 12 groups overlapping) into user's libraries by collapsing the energy group structure of 53 group 2 fas(2 st 1 thermagroup3 d an sl groupscele th l r calculatio)fo f no the fuel element. We selected the thermal cut off energy at 1.12 eV so thalow-lyine th t g Pu-240 resonance leve treates i l d with up-scattering effect. Cell calculation and collapsing the energy structure We performed the cell calculation by the collision probability method throug whole th h e energy range Bondarenka . V e Abov 0 13 e o type tabl r resonancfo e e shielding factor e effectivth uses r i s fo d e cross sections A fixe. d source proble9 5 spatiagroup1 1 d r an sfo ml regions i s solved for this fast energy range. Between 130 eV and 1.12 eV, about 5000 linear equations corresponding to 5000 energy points for 19 spatial regions are successively solved for rigorous resonance integral. In thermal energy rang fixea e d source proble5 spatia1 group2 3 d r an sfo ml regions is solved. After the above cell calculation the energy group structure of cross sections is collapsed from 53 to 10 groups (3 groups are also examined) for two dimensional diffusion core calculation. The energy structure use shows e geometri d Th Tabln i n . 1 e y assume shows i d n in Fig. 1.

232 Burn-up process The cell burn-up process follows at each end of cell calculation. The chain scheme currently used is shown in Table 2. Results K-infinite's are shown in Table 3. The chang.es of atom number densities during burn-u e showar n p i Tabl b a n'LE 4- 4- 'HEU eU. d an < K-effective' e show ar sn Tabl i n 5 etogethe r wite result th h3 grou y b sp structure. Reactivity coefficient e showar sn Tabl i n . Absorptio6 e n reaction rate f specifieo s d isotope d theian s r ratio e comparear s n i d Table 7. Fig.l Cell Geometry for Collision Probability Calculation

~~\ \ \~~

Radius Ar T* R** Material * Radius Ar T R Material * •* M X X X x-i:x^-Jr.-*.-Jfir. A'^^fJ^yf. •JfSf^-jfJf^fJkJfx::•*:-jr.xv.-J{.*M*.>:^>:.->tv.'Jr.-Jfjf^W-k v. s.*.*}:-**:-»: jf.Jf.TKV. Jr. 1 .013 1 .013 1 \ D20 * 4.171 051 14 12 Al 2.026 1.013 2 \ D20 > 4.509 338 15 13 D20 3.039 1 .013 3 \ D20 •* 4 .560 051 16 1 4 Al 3.090 .051 4 2 Al x 4.610 050 17 15 Méat 3.140 .050 5 3 Méat -• 4 .661 051 18 16 Al 3. 191 .051 6 4 Al x 4.999 338 19 1 7 D20 3.529 .338 7 5 D20 ' 5.156 157 20 18 Al 3.580 .051 a 6 Al r 5.844 688 21 1 9 D20 3.630 .050 9 7 Méat ,. 6.531 687 22 19 020 3.681 . .051 10 8 Al •* 7 .218 687 23 1 9 D20 4.019 .338 11 9 D20 >- 7 .905 687 24 1 g D20 4.070 .051 12 10 Al o .598 693 25 1 g 020 4.120 .050 13 11 Méat * *• ...... • • • ...... •:.:> .K ~Jt x x. x :*::»• ^~** :T' x ^ :». >î «• »*•*•' M •- '•'' - '•' '' '•'' •• • ' ' * • • =' - " ' ' ' T- Spatial sub-division for cell calculation in thermal range R«. Spatial sub-divin i n s io resonance range

233 Table 1 Energy Group Structur n DIDi e O type Benchmark Calculations 53 G Upper 10 G 3 G Remarks Number Energy Number Number *******>fC3fC3fC9fC^CJ|ClfC^C^C3fC^C: |t**##*#s| 1 10.00 HeY l 2 6.065 l 3 3.679 l 4 2.231 l 5 1 .353 2 6 820.V 9Ke 2 7 497.9 2 8 302.0 2 9 111.1 3 10 7 48 0. 3 l 11 15.03 3 l 12 5.531 3 l 13 2.035 4 l 14 748.V 5e 4 1 15 353.6 4 2 16 130.8 4 2 Cut off for table-look-up 17 78.89 4 2 18 37.27 5 2 19 17.60 5 2 20 8.315 5 2 21 3.928 5 2 22 1 .855 5 2 23 1.125 6 3 Thermal cut off energy 24 .6825 6 3 25 .4140 6 3 26 365.V 3me 6 3 27 319.6 6 3 28 297.9 6 3 29 277.0 6 3 30 237.4 7 3 31 200.9 7 3 32 183.8 7 3 33 167.4 7 3 34 151 .8 7 3 35 137.0 7 3 36 122.9 8 3 37 109.6 8 3 38 97.08 8 3 39 85.40 8 3 40 74.28 8 3 41 64.02 8 3 42 54.52 9 3 43 45.79 9 3 44 • 37.81 9 3 45 30.60 9 3 46 24.15 9 3 47 18.47 9 3 48 13.54 10 3 49 9.381 10 3 50 5.980 10 3 5t 3.342 10 3 52 1 .466 10 3 53 .3524 10 3

234 Tabl 2 e Burn-up Chain Scheme use DIDO-typn i d e Benchmark Calculations Chain scheme taken from ENDF/B II Nuclid) 1 e Characteristics ZZM Fissile Decay const Power/fiss U05 1 0.0 3.10800E-11 U08 1 0.0 3.I0800E-11 PU9 1 0.0 3.22000E-11 PUO 1 0.0 3.22000E-11 PU1 1 l.68000E-09 3.22000E-11 PU2 1 0.0 3.22000E-11 XE5 0 2.11000E-05 0.0 SM9 0 0.0 0.0 F5N 0 0.0 0.0 non-saturated F.P. F5S 0 6.00000E-10 0.0 slow-saturated F.P. F5R 0 2.10000E-10 0.0 rapid-saturated F.P. F9N 0 0.0 0.0 non-saturated F.P. F9S 0 4.40000E-10 0.0 slow-saturated F.P. F9R 0 1.66000E-10 0.0 rapid-saturated F.P.

Heav) 2 y Nuclide Chains U05 U08+PU9+PUO+PU1+PU2 3) Absorber Chains none 4) F.P. Chains F5N.F5S.F5R from fission of U05 F9N.F9S.F9R from fission of PU9 5) Fission Yields U05 U08 PU9 PUO PU1. PU2 XE5 0.066 0.065 0.075 0.073 0.072 0.073 SM9 0.011 0.018 0.013 0.024 0.015 0.024 F5N 1 . 5403 0.0 0.0 0.0 0.0 0.0 F5S 0.378 0.0 0.0 0.0 0.0 0.0 F5R 004. 0 7 0.0 0.0 0.0 0.0 0.0 F9N 0. 0.0 1.507 0.0 0.0 0.0 F9S 0. 0.0 0.394 0.0 0.0 0.0 F9R 0. 0.0 0.011 0.0 0.0 0.0

235 Table 3 K-infinity of the Fuel Element *U-235 g burnt* HEU LEU * 0 * 1 .8752 1 . 7688 * * 20 * 1 .7335 1 .6382 * * 30 * 1 .7126 1.619t , * * 40 * 1 .6889 1 .5981 * * 45 * 1 .6766 587. 1 3 * * 60 * 1 .6312 1 . 5492 *

~~Table 4-a Number Densities for HEU Fuel U-235 g burnt Xe-135 Sra-149 U-235~~

***##*#*#*5f** •T»T»*r"T-*T»T»'I*»l>*T"l»*T»T1>n 0 0.0 0. 302E-0. 1 3 20 1 . 269E-08 9.586E-08 1 .128E-03 30 1.171E-08 8.819E-08 1 .042E-03 40 1 .082E-08 8. 109E-08 9.548E-04 45 1.029E-08 7.720E-08 9.1 14E-04 60 9.Q77E-09 6.733E-08 7.813E-04

~~U-235 g burnt Pu-239 Pu-240 Pu-241 :**********:(:***********#**** !f:5jc5fe;ie5jC5ic>:5;C3j;>cx:*i>-i>O >rU •st• -•! <~~

~~Table 4-b Number Densities for LEU Fuel U-23 5burng t Xe-135 Sra-149 U-235 U-238 :**************»fC^c5fC>r>jcxxx-x-cx^x^- K U-23 burn5g t Pu-239 Pu-240 Pu-241 Pu-242~~

0 0.0 0. 0. J*. 20 1.051E-05 4.728E-07 4.031E-08 7.966E-10 X 30 1.499E-05 028E-06 1 .328E-07 4.225E-09 «« 40 1.887E-05 750E-06 3.004E-07 1.369E-03 - 45 2.057E-05 164E-06 144E-0. 4 7 2.214E-C8 • 60 2.501E-05 3.579E-06 9.047E-07 7. 144E-08 -

~~236 Tabl 5 e K-effectiv r DIDfo eO type Benchmark Problem HEU LEU IOC 3G 10 G 3G~~

~~• Fresh 1 . 3092 1.3139 2741. 2 1 . 2786 - BOL -4 8 74 1 1. 1.1782 1. 146 3 1 . U81 • EOL 2 43 1 1. 1.1432 1. 121 3 1.1226 • 4Û. g burnt» 1 .1555 1 . 1 593 1 . 1308 1 . 1308~~

Table 6 Feedback Coefficients for U-235 40 g burnt Core (ApSO HEU LEU=- • D20 Temp 20 c - 50 c .401 -.398 •• Fuel Temp 20 c - 300 c .001 -.381 • 20*. Void in 2 element .287 -.275 x A^ • • ,*.* « ,*:«•. A •.*•*. - M.» A^. ^x^xcA^c

~~Tabl e7 Absorption Reaction Rate Specifier fo s d Isotopes~~

~~U235 Mo98 B 10 Co59 N 14 •.**•**.** r •<.*. Core 93» . 537+8 . 667-5 .306+9 .378+7 .154+6 Center 20*i .458-8 .653+5 .26U9 .312+7 .131-6 (20/93 > .853 .978 .851 .872 .851~~

~~ï^MCX^cAx^cifC^cskiic^c: Ä 93 76mm .495+8 .234+5 .285+9 .298+7 . 141-6 into Reflector 20Ä .465-r8 .233+5 . 268-9 .282+7 . 133+6 «20/93) .941 .996 .941 .944 .941~~

~~237 APPENDI5 F- X~~

~~DANISH BENCHMARK CALCULATIONR FO S HEAVY WATER RESEARCH REACTORS~~

~~C.F. H«5jerup~~

~~National Laboratory Departmen f Energo t y Technology~~

~~239 1 . Introduction~~

Calculation e IAEth A r benchmarfo s k problem have been performed using programs developed at Rise. In the course of the benchmark a numbe f modificationo r e programth o t s s have been made, minor errors detected and corrected, and output editing changed to serve specific purposes.

e biggesTh t changee programth n i s s were introductioth e a f o n new method for resonance calculations (the socalled sub-group method (1) (2)) necessitated by the complex annular geometry of the benchmark fuel elemente generatioth w d resonancne an , f o n e data libraries for 93% and 20% enriched fuel.

~~. Method2 f calculationo s s~~

~~Cell calculations with burn up were performed with the program CCC (3), modified as described in the introduction.~~

This program calculates by collision probability methods a 76 group flux spectru a geometr n i m y somewhat simpler thae th n actual one (only 10 regions). 10 group cross sections for all nuclides are condensed from the 76 group library. Thermal scattering cross sections are generated by a modified Nelkin metho r deuteriufo d d hydrogenan m l a fres al mode d ga r ean ,fo l other nuclides.

~~e resonancTh e cross sectio r U235fo n , d U23Pu23e an 8ar 9 calculate a sub-grou y b d p method using collision probabilities e spatiafoth r l effects.~~

0 grou1 e p Th energy boundaries usee giver ar d Fo Tabl n i n . 1 e the subsequent diffusion calculations cross sections of the homogenized fuel element in 3 energy groups are generated and stored on a file.

~~As a new output feature the 3-group cross sections and reaction rates of all the nuclides were orinted.~~

~~240 e globaTh l calculation y geometrx- n i s y were made wit a difh - fusion theory programme, capable, as well, to do perturbation theory estimates of reactivity effects.~~

~~Table 1~~

~~Energy boundaries in.10 groups and 3 groups (Lower energ = 0.000y ) eV 1~~

~~upper 15 0.498 5.531 748 1 .855 1 .099 1 .014 0.625 0.251 0.043 energy MeV MeV keV eV eV eV eV eV eV eV~~

~~Group 1 2 3 4 5 6 7 8 9 10 (10) Sroup 1 2 3 (3)~~

~~3. Results~~

~~Cell burn up calculations The k« values from the cell burn up calculations are given in Table 2 for the two enrichments.~~

~~Reaction rate distributio n nuclideo n t zera s opres i bur -p u n sente d concentration Tabln an i d , 3 e f uraniuo s d plutoniuan m m isotopes, Xe135 and Sm149 are given in Tables 4 and 5.~~

~~Overall diffusion calculations The results from the x-y diffusion calculations are presented e nexith n t tables.~~

~~In Table 6, the keff values at the BOG and EOC for the two enrichments are given.~~

241 The 3 group fluxes along the x-axis at the core midplane for the 93% BOC case are listed in Table 7. The normalization is such that the power per cm length of the z-axis is 10 MW/61 cm, assuming 3.1*10^ fission per Joule. If an axial sine distribution with zeroes at 22.7 cm from top and bottom is assumed, the value n Tabli s 7 shoule e multiplieb d y 1.14b d o yiel9t. d peak values.

~~C fluxe'BO e % ar s93 e C fluxeth BO o % t s20 e e ratioth Th f o s give n Tabli n . Bot8 e normalize ar h o yiel t de sam th de total power of 10 MW.~~

~~Reactivity coefficients calculations~~

e reactivitTh y e specifievalueth f o s d changes {channel temperature, fuel plate temperature, and coolant density) were obtained by perturbation calculations. The modified diffusion cross sections were generated at a burn up of 40 gr U235 per element and the x-y calculation set up with a uniform core of that bur-n up. The results are given in Table 9. The diffusion coefficient e modifie e th bas th r er fo d casfo s d casean e e ar s given in Table 10.

~~Calculation of reaction rates of different isotopes at centre f reactot x=53.o a d m c an 2r~~

~~Only very few of the isotopes mentioned at the Vienna meeting are availabl libraryr Ud 2ou an ^ n onls . i i n efac^ I B .yt i t~~

r theso Fo isotopetw e e infinitth s e dilution reaction "rate ratios between 20% and 93% cores with uniform burn up to 40 gr. U235/element have been calculate e spectrth n i da e founth n i d centre of the core and in the reflector position 53.2 cm along the x-axis e resultTh . e givear s Tabln i n. 11 e

~~242 References~~

~~M.J) (1 . ROTH. Resonance Absorptio Complicaten i n d Geometries AEEW-R921 , 1974.~~

~~(2) H. NELTRUP. Private communication.~~

~~(3) C.F. Hajerup. The Cluster Burn up Programme CCC and a Com- parison of its results with NPD Experiments. Risa-M-1898, 1976.~~

~~Tabl2 e a functio s a « f buro n p u n~~

~~93% enrichment 20% enrichment~~

~~g-rams U235 grams U23^ burnt k« burnt k«~~

~~0 1 .874 0 1 .764 1 .0 1 .786 1 .0 1 .683 10.0 1 .756 10.0 1.655 20.0 1 .733 20.0 1 .636 30.0 1 .712 30.0 1.617 40.0 1 .688 40.0 1 .597 50.0 1 .661 50.0 1 .574 60.0 1 .629 60.0 1.549~~

~~243 Tabl3 e~~

~~Reaction rates at Zero burn up in k« lattice calculations~~

~~93% enrichment % enrichmen20 t Nuclide Group fission capture Fission capture~~

% 1 0.00288 0.00066 0.00316 0.00073 U235 2 0.04146 0.02216 0.04263 0.02257 3 0.72677 0.12813 0.67881 0.11984 Total 0.77110 0.1 5096 0.72459 0.14313 1 0.00002 0.00003 0.00104 0.00179 U235 2 0 0.00263 0 0.05880 3 0 0.00027 0 0.01339 Total 0.00002 0.00295 0.00104 0.07502 1 0.00124 0.00118 Al 2 0.00512 0.00467 3 0.05863 0.04782 Total 0.06499 0.05367 1 0.00335 0.00333 D20 2 0.00024 0.00023 3 0.00313 0.00269 Total 0.00672 0.00625 1 0.00001 0.00001 »2° 2 0.00014 ' 0.00012 3 0.00339 0.00239 Total 0.00354 0.00252

~~244 Table 4~~

~~Number densities for 93% fuel (Nuclei per 10~24cm3)~~

g U235 Xe135 Sm149 U235 U236 U238 burnt 0 0 0 1 .302-3 0 9.679-5 20 1 .308-8 1 .021-7 1 .129-3 2.826-5 9.619-5 30 1 .219-8 9.891-8 1 .043-3 4.222-5 9.588-5 40 1 .128-8 9.560-8 9.564-4 5.604-5 9.556-5 50 1 .037-8 9.203-8 8.703-4 6.973-5 9.523-5 60 9.452-9 8.788-8 7.843-4 8.328-5 9.489-5

g u235 Pu239 Pu240 Pu241 Pu242 burnt j 0 0 0 0 o. 20 4.707-7 2.121-8 1 .815-9 . 4.140-11 30 6.796-7 4.807-8- 6.237-9 2.329-10 40 8.534-7 8.343-8 1 .442-3 7.820-10 50 9.933-7 1 .256-7 2.691-8 1 .988-9 60 1 .100-6 1 .729-7' 4.386-8 4.262-9

~~245 Tabl5 e~~

~~-Number densitie % fue20 lr (Nuclefo s r 10~pe i 24cm3}~~

g CJ235 Xe135 Sm9 14 U235 U236 U238 burnt 0 0 0 1 .454-3 0 5.744-3 20 1 .483-8 1 .163-7 1 .279-3 2.890-5 5.729-3 30 1 .407-8 1 .143-7 1 .192-3 4.300-5 5.721-3 40 1 .329-8 1 .122-7 1 .107-3 5.686-5 5.713-3 50 1 .250-8 1 .099-7 1 .022-3 7.046-5 5.704-3 60 1 .170-8 1 .070-7 9.385-4 8.381-5 5.696-3

g U235 Pu239 Pu240 Pu241 Pu242 burnt 0 0 0 0 0 20 1 .203-5 5.047-7 4.216-8 8.771-10 30 1 .751-5 1 .138-6 1 .438-7 4.830-9 40 2.223-5 1 .970-6 3.305-7 1 .587-8 50 2.624-5 2.958-6 6.128-7 3.940-8 60 2.956-5 4.074-6 9.940-7 8.242-3

~~Table 6~~

~~keff from global x-y-calculations~~

~~Enrichment Description keff Ap%~~

~~BOC 1 .1668 93% EOC 0 34 1 . 1 2.48 BOC 1 d433 20% EOC 1 .1177 2.00~~

~~246 Tabl7 e~~

~~93% enrichment. BOG fluxes at core midplane~~

~~x(cm) ' 2.53 22.8 32.93 38. I 43.00 7 14 *1 0.690 0.628 0.567 0.504 0.394 x10 14 *2 0.882 0.789 0.666 0.573 0.455 x10 4>3 1 .332 1 .173 1 .021 0.969 0.970 x10 14~~

~~x(cm) 48.13 53.2 58.27 68.4 83.6 88.67 1 .782 0,795 0.354 0.0702 0.00631 x10 13 *1 0.00304 13 4>2 3.189 2.017 1.212 0.413 0.158 0.203 x10 13 *7 10.825 10.620 9.718 7.301 4.103 3.212 x10~~

~~x (cm) 93.7 ' 3 98.8 103.37 114.0 ' 1 21 .6 *' 0.0181 0.0105 0.00607 0.00171 0 12 *2 3.571 3.701 3.139 1 .402 0 x10 12 *i 23.237 16.984 12.203 4.790 0 x10~~

~~The fluxe e normalizeac s d such lengta thae er powe th r th pe r along the z-axis is 10 MW/61 cm assuming 3.1'10^° fissions/Joule.~~

~~247 Tabl8 e~~

~~% enrichmenRati20 f o t % enrichmenfluxe93 o t s t fluxes along x-axis at core midplane.~~

~~x(cm) 2.53 22.8 32.93 38.0 43.07 RI 1 .032 1 .028 1.017 1 .018 1 .023~~

~~R2 1 .005 1 .001 1 .000 1 .001 1 .003 R-? 0.862 0.868 0.880 0.890 0.905~~

~~x(cm) 48.1 3 53.2 58.27 68.4 83.6 88.67 RI 1 .023 1 .025 1 .026 1.027 ' 1 .028 1 .028~~

~~R2 1 .008 1 .011 1 .013 1 .014 0.988 0.976 R-î 0.933 0.947 0.954 0.962 0.965 0.966~~

~~) cm { x 93.73 98.8 103.87 1 1 4.0 ! 1 21 .6 Ri 1 .028 1 .028 1 .028 1 .028 - 0.969 0.967 0.967 0.967 - R2 R? 0.966 0.966 0.966 0.966 -~~

~~248 Tabl9 e~~

~~Reactivity coefficient U23g 0 5 (4 sburnt )~~

~~Case Fuel description type k. keff Ap% Âp (slope)~~

~~93% 1 .68810 1 .1479 - - Base 20% 1 .59660 1 .1266 •• "•~~

~~50°C coolant 93% 1 .68608 -0.488 -1.63lQ-4 and fuel 20% 1 .59363 -0.457 -1 .5210-4~~

~~300°C fuel 93% 1 .68774 -0.045 -1 ,6010-6~~

~~20% 1 .58708 -0.448 -1 .6010-5~~

~~20% coolant void 93% 1 .68591 -0.315 -1.57lo-4 (centra elem2 l .) 20% 1 .59268 -0.296 -1 .48^-4~~

~~Table 10~~

~~Diffusion coefficients~~

~~Case 93% 20% description Group 1 2 3 1 2 3~~

~~Base 1 .3692 1 .2937 0.9149 1 .3666 1 .2834 0.9140~~

~~50°C coolanh c t 1 .3730 1 .2968 0.9184 1 .3705 1 .2864 0.9171~~

~~300°C fuel plate .3691 si 1 1.2937 0.9151 1 .3660 1 .2829 0.9140~~

~~% coolan20 t void 1 .4711 1 .3734 0.9683 1 .4632 1 .3615 0.9661~~

~~D2O reflector 1 .3085 1 .2223 0.8775~~

~~Graohi te ' 1 .1884 0.8660 0.8618~~

~~249 Table 11~~

~~Reactio Ud 23an ^ n(onl^ 8 rate c y fo scapture cor2 n ei ) positions for 20% and 93% cores, uniformly burned to 40g~~

* *B (jcoB c235

Gr.1 1 .8496 2.2274 4.12 0.45000 0.83 0 Gr. 2 2.396 165.34 383.52 13.839 32.10 Gr. 3 3.4071 2845.4 9694.56 72.529 247.11 93% Total 10082.20 280.05 Gr.1 0.18134 2.2274 0.40 0.45000 0.08 Gr. 2 0.48630 165.34 80.40 13.839 6.73 53.2 Gr. 3 2.6755 2845.4 7612.87 72.529 194.05 Total 7693.68 200.86 Gr.1 1 .8618 2.2274 4.15 0.45000 0.34 0 Gr. 2 2.2864 161.45 369.14 13.290 30.39 Gr. 3 2.9055 2790.1 81 06.64 71 .055 206.45 20% Total 8479.92 237.67 Gr.1 0.19596 2.2274 0.44 0.45000 0.09 Gr. 2 0.49471 165 1.4 79.38 13.290 6.57 53.2 Gr. 3 2.5106 2790.1 7004.83 71 .055 178.39 Total 7085.15 185.05

~~«20% (B10/ 0} = 0.841 (B10, 53.2} » 0.921 »93%~~

~~(U235, 0) » 0.849 (U235, 53.2) = 0.921 »93% R93%~~

~~$(20%) (0) = 0.853 (53.2) 0.938 $(93%)~~

~~250 APPENDIX G~~

~~List of Participants~~

~~r" ———— — —t— ———— —• - — -p _£"u) i pt J-H t 01 cd -p CQ S >•> D cd C\J 0 00 H CQ r-i ÏS C\ iH 00 H CO CO cc 1 O\ 1 O^ CO ON " O C —~~

~~ADDRESS~~

~~AUSTRALIA i D.B. McCulloch X X X~~

~~AAEC ! Sutherland, N.S.W. 1~~

CANADA D. Axford X Atomic Energy Canada Ltd. Chalk. River Nuclear Laboratories Chalk River, Ontario l M. Feraday X 1 Atomic Energy Canada Ltd. 1 Chalk. River Nuclear Laboratories 1 Chalk. River, Ontario

~~Garve. P y X X X I Atomic Energy Canada Ltd. Chalk River Nuclear Laboratories i Chalk River, Ontario 1~~

~~DENMARK~~

~~K. Haack X X A X Riso National Laboratory DK-4000 Roskilde~~

~~C.P. Hoyerup X X X X Riso National Laboratory ' \ DK-4000 Roskilde~~

FRANCE F. Joly X CERCA 9-11 rue Georges Enesco 1 94008 Creteil Cedex \1 1 GERMANY FED. REP. K-J. Kalker x X X Zentralabteilung Forschungsreaktoren Kernforschungsanlage Julien Postfach 1913 D-5170 Jülich

251 i' ü v cd •p 02 ^ ft ^a 03 CM C ro H CQ H S OJ H OC H co 00 co A O C 1 1 v co O\ t— OA H r- CM H H 1 H 1 H H t— MD GERMANY FED. REP. (Contd.) H-J. Regler X X X INTERATOM Friedcich-Ebert-Strasse D-5060 Bergisch Gladbach l

~~Schuet. ü t X NUKEM GmbH Postfach 110080 D-6450 Hana1 u1~~

~~JAPAN S. Matsuura X Offic f Plannino e g Japan Atomic Energy Research Institute Fukoku Seimei Bldg2 .2- Uchisaiwai-cho-2 chôme Chiyoda-ku, Toky0 o10~~

~~K. Kanda X X X Research Reactor Institute Kyoto University Kumatori-cho, Sennan-gun Osaka 590-04~~

K. Tsuchihashi X X X Office of Planning Japan Atomic Energy Research Institute Fukoku Seimei Bldg. 2-2, Uchisaiwai-cho 2 chôme Chiyoda-ku, Tokyo 100 U.K. Constantin. G e X X X X AERE Harwell Materials Physics Division Bldg. 775 Oxfordshire 0X1A 1OR

~~R. Panter X X X X AERE Harwell Research Reactors Division Bldg. 521 Oxfordshire 0X11 ORA~~

252 -p o ft -P CQ 1 a 0l œ CM O on H en H S CM c-i CO H CO co CO 1 ON N O 1 CO ON C- ON i-l r-t CM H 1 iH l r- 1 1 — i i-l c— ^Û U.S.A. A. Travelli X X X Argonne National Laboratory 9700 S. Cass Avenue Argonne, 111. 60439

~~J. Hatos X X X X Argonne National Laboratory 9700 S. Cass Avenue Argonne, 111. 60439~~

~~Hoeni. M g X Arms Contro Disarmamend an l t Agency 20451 Washington D.C.~~

~~YUGOSLAVIA R. Martine x Boris Kidric Institute of Nuclear Science 11001 Beograd~~

~~Scientific Secretary R.G. Muranaka~~

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