maflan of a prehmmary ‘yse at the Ok Ridge Nattonu{ r correchon ond ?herefere cfoe' T LEGAL'RO?IEIE.: - S L This report was prepored as an account of Gevernmenf sponsoted work "Neither the Umted Stutes, nor the Commission, nor any person acting on behalf of the Commission: ' , T " A, Makes any warronty -or. ‘représentation,’ .xprassed -or implied, with respect to ihc ‘accuracy, - completeness, or usefulness of the information contained in this report, or that the use of - privately owned rights; or -- o e ) ony informetion, apparatus, method or process disclosed in this report, ol .7} ---As used in the above, ''person acting on behali of the Commission® includes any umployee or =+ 1° contractor of the Commission, or employee of sweh ‘Eontractor, to the extent that such empfoyee - 1. or contractor of the Commtssnon, or’ cmp!oyee of such contracter prepares, disseminotes, -or ~ provides access to, any information’ pursvont fo hls cmp!oym-m of contract with the Commusslon, " or his employment with such :onfroctor. ) - - ~ any information, apporatus, method ot - proc-ss dnsclosed in fins report ‘may not mfrmge B. Assumes any liabilities with fospcct to the use c! or for damuges nsulhng from the use of e - g v , : - ORNL TM-1060 Contract No. W-7405-eng-26 5 REACTOR DIVISION & f MOLTEN SALT CONVERTER REACTOR Design Study and Power Cost Estimates for a 1000 Mwe Station L. G. Alexander, W. L. Carter, C. W. Craven, D. B. Janney, T. W. Kerlin, and R. Van Winkle N jp SEPTEMBER 1965 OAK RIDGE NATIONAL ILABORATORY Oak Ridge, Tennessee - Operated by UNION CARBIDE CORPORATION for the U. S. ATOMIC ENERGY COMMISSION »: o f o e W (® 4! w‘ ° < AT— Hcéfiw(jqimmmew iii FOREWORD A molten-salt-breeder reactor was evaluated at the Oak Ridge National Laboratory beginning in 1959. Because a number of the features postulated had not been demonstrated at that time, the realization of a breeder ap- peared to lie rather far in the futuie. Accordingly, the study of the near-term, one-region, one-fluid molten-salt converter described in this report was begun in July 1961 and completed in December 1962. Since then, several advances have been made in molten-salt technology which make the breeder reactor much less remote and modify some of the conclusions in this report. Briefly, these advances include: 1. Progress in core graphite design which greatly simplifies previ- ous problems of‘separating the oore into two regions — one for the uranium- bearing fuel salt and one for the thorium-bearing blanket salt. The new design utilizes a liquid-lead seal around the tops of graphite tubes con- taining fuel salt that allows the tubes to expand or contract freely while maintaining an absolute seal between fuel and blanket fluids.¥ The addi- tion of a blanket results in a much better conversion than obtained in this report and leads directly to an attractive breeder. 2. Thermal engineering studies which show that the Loeffler boiler system can advantageously be replaced by a supercritical boiler. Thermal stress problems sre reduced, overall thermodynamic efficiency is increased, and capital costs are considerably reduced. In addition, studies of so- dium metal and of mlxtures of alka11 carbonates show that if either of _these 1nexpensxve materlals can be safely used for the intermeédiate cool- ant in place of the costly 11th1um-bery111um fluoride mixtures postulated in this study, then further large cost reductions can be realizedj' ~ *E. 8. Bettis, Oak Rldge Natlonal Laboratory, personal communication -w1th L. G. Alexander, Oak Rldge National Laboratory, January 1965 TC W. Collins, Oak Rldge Natlonal Labo*atory, personal communlcatlon w1th L. G. Alexander, Oak Rldge Natlonal Laboratory, January 1965 ‘iv 3. A fuel purification process based on simple distillatiomizwhich_' not'only reduces processing costs§ but permits reuse of the carrier salts — an advantage not assumed in this study. As a result of these developments, we believe that fuel cycle costs for a two-region breeder based on 1965 technology will be only 0.3 to 0.4 mill** compared to the 0.68 mill/kwhr shown in Table 6.10 for the MSCR. TM; J. Kelley, Oak Ridge National Laboratory, personal communication with L. G. Alexander, Oak Ridge National Laboratory, January 1965. $W. L. carter, Oak Ridge National Laboratory, personal communication with ‘L. G. Alexander, Oask Ridge National Laboratory, January 1965. **H. F. Bauman, Osk Ridge National Laboratory, personal communication with L. G. Alexander, Oak Ridge National Laboratory, January 1965. “ g~ " A O - - ] ;g s v ”‘fij | _ - CONTENTS Page ABS AT e iiiitiitiieiereaetennnneanssenssnannnnns tresenenn Cecoanea 1 1. SUMMARY «ovvvnnnvnnessnn. e eeearreae ettt rereaaas R | j 1.1 Description e, cessnee ceannn Cetes et steannannans 3 § 1.2 TFuel ReprocesSSing seeeesevneeceneennnns Chesecesesscsanssanens 3 § | 1.3 Nuclear and Thermal PerfOIMANCE «vevevernernesnnenss Ceeenaas 3 % 1.4 Fuel Cycle Cosfi e ceeras 4 é - 1.5 Power Costs ...... Cecsaanns S e e e e esssesas e tettnnanan Cheeeas 5 B 1.6 Advanced MSCR +.......... e, ettt i, 6 5 " 1.7 Post Script = JanUATrY 1965 1 ivtteetententcnnsenecnnonnannnn. 7 § 2. INTRODUCTION ..... Ceeeesens Cesesssesennraans Ceeenenn et erteaaae 8 g 2.1 Purpose, Scope, and Method of Approach ceseeee e v 8 2.1.1 Figure of Merit -veeeveveeeeesennns e, : 8 § 2.1.2 Reactor Concept ..e.vvveeenennn Cetetesrtiasecnenne veee. 8 é 2.1.3 Procedure seeeeaans et eecessaat st eretestneseanenas e 8 E 2.2 Status of Molten Salt Reactor Development -........ R . 9 2:.2.1 Early Work ...c....... csesresenans vessssasenan tevansae 9 % . 2.2.2 The Molten Salt Reactor Program ....eeeeveeves.. cienes 10 g 2.2.3 Fuel Development .....vevveeevenen. beseenaceas ceevsse. 10 é 2.2.4 Container Development ..c.cvvvennnn. teseses Pereeaaaase 12 § ' 2.2.5 Moderator Development........, ....................... 13 ééi 2.2.6 Component Developmeet .i....;...........f;.....f...... 14 | - 2.2.7 V_Reactor Vessel_,..;;,,...,.g........,................. 14 - 2.2.8 Molten Salt PUMPS +vvr v, Crrierieeiiisieenes 14 2.2.9 ~ Molten Salt Heat Exchangers and Steam 3011ers ceseees. 14 . 72.2Ql0'_Freeze Valves and’ Freeze FLANZES +»rvevevererannnnennns 15 f2{2.11 Molten Salt Instrumentatlon and Special o . Emummnt.q;g.p.“,. ...... “.”.”.g.n.“.g.“. 16 7212§12 Remote Maintenance l...;.{....;;..;...;,;.5;.;.... 16 72.2.13_ Chemical Proce851ng of Molten Salt Fuels +veveveerea.. 18 L 2.2.14 Fluoride Volatlllty and HF Solution Processes .ve..... 18 5 ‘fi; . 2.2.15 Thorex ProCesSs +veeeseennencoronennnns Creeeaanan v 19 > ¥ vi CONTENTS (continued) | 2.2.16 Fractional Crystallization Process .«....-. eeaeeaaana 2.2.17 dther Processes ..csceveenans Ceesteateeseaseitteanans 2.2.18 Molten Salt Reactor Studies ...covevuvvnnns Cessersasas 2.3 Molten Salt Reactor Experiment ..... cereen e rrere e .. 3. BASES AND ASSUMPTIONS i e eereeeeeeneeae 3.1 Design Bases ....-. vecssannn et tseesarsoaanass ceraan ceaenenn '3.1.1 Reactor Concept +.covevrencuicerinnennnaen S P 3.1.2 Design Calculations ...... esesesanacscstasbanns “ens 3.1.3 Station Power .....cieiieeeaen. eeeas 3.1.4 Plant Utilization Factor ....... reesvasisenna ceeanans 3.1.5 Thermal Efficiency eceeevececocenanss Cerresaereenenonn 3.1.6 Fueling Cycle «seecesncnass ceeanans tresesune toieesesane © 3.1.7 Processing «ieiceiiiiiiiians tessesseas vevesnes tevsans 3.1.8 Feed and ReCYCle «eeeieriersacnnsossnsnnssonsnsasenasns 3.1.9 Isotopic Composition of Lithium ......... R tesece 3.1.10 Energy Conversion System ......cconve teeaccasassensos . 3.1.11 Primary Heat Exchanger Requirements «.....eceeeccacas ' 3.1.12 Minimum Salt Temperatures ...ececeeceee. ceeaen ceesans 3.2 COSt BASES cvevencscrocancssans PO Cerereeesreareacana s 3.2.1 Value of Fissile Isotopes «ccssesscccccsascses ceerena 3.2.2 Value of Thorium ....... et isesesseacsensntiossvannan 3.2.3 Value of LiF(99.995% 7Ii) eevrveevennennn 3.2.4 Value of BeFg «..v..n teseesasansaans cseaceensacsasiaces 3.2.5 Value OFf Base S81L «evveneetnoeennaaseasnoanens 3.2.6 Cost of Compounding é.nd Purifying Fuel Salt ..coveevs 3.2.7 “INOR-8 COSt cevenvcrsnnncanonseranas Cesesesecenne Ceves 3.2.8 Moderator Graphite COSt «eeeecesoennaeoreansannnesans 3-209 AmualFixedcharges OOOCOD"l_-._..‘l....,...c...q.ll.. 3.2.10 Central Fluoride Volatility Plant Processing Charges .ceeeeeseess Ceeeetentertetetet it 3.3 Special Assumptions'.;.. ..... s etesaatasestisastesnann cen 3.3.1 Permeation of Graphite by Salt .......... seessesesaes Q) Xy Ny o Y * !'(p AT A “ % “ M‘gdejasflMw_,"mb MWMMW_MWwWWwWMMNMWWWW o U ‘i} vii CONTENTS (continued) 3.3.2 Permeation of Graphite by 2®Xenon ......ievviiniiian. 3.3.3 COrrOSion PrOQUCHS « o vt onntnrernennsnneenennennennss 3.3.4 Approach to Equilibrium ........ e Chresaee e .. DESCRIPTION OF MSCR CONCEPT «vvveeviecnnevonnnans Cenasenas veo 4.1 General Description eeeeieeriiericaanansrneenans beeesenann 4.2 Site Plan «evesesrrcnnsssosarancans veseens Ceeseseane teaenan 4.3 Structures .......n . Netsesaraasaaasnn 4ol Primary System COmponents ..cieeeesnsesvnocecronsssnsennsans 4eteel Reactor VeSSel +1.veieieessnctoossstsanrsnssnssaannons 4.4.2 Moderator Structure ......co i, Cererreaaea 4.4.3 Fuel-Salt Circulating Pumps .eeceersvoccnnccaanecnan, 4.b.4 Primary Heat Exchanger ........ccviecieiinenannn .o 4.5 Intermediate Cooling System .......coveiaciannn vesasa caeees 4.5.1 Introduction ............ vesaes Ceseeaans theresseanara 4.5.2 COOLANEL SALE PUIMDS. « v reenrennsnnnennrnneeneennennes ) 4.5.3 Steam Superheaters «..eeveeecaes beeeae Peerseisenaanns 4.5.4 Steam Reheaters +eveerverecsscasasroansnn ceeeas cieen. 4.6 Power Generation SyStem +c.ceeiiiiiienistoirtrerarsessasanns 4.6.1 Introduction ........ Peease et aas teeresiearann 4.6.2 Loeffler Boiler System ....... Cerenean beeseaserenssan 4.6.3 Steam Circulators «sec.ev.oo.. Ceennsan Cees v bebod 'Turbogenerétbr g;.;.f..Q....., ..... Ceerrseenen vereaas 4.7 Reactor Control Sysfiéfi-,.w,,..,.... ...... Cecenees cevesannas 40721 INBLOQUCHION +oveersvrenrnnennerensaseneneneenennen. 4.7.2 Shim CONtrOl «evevvvsseesnsnnnnns R P - 4.7.3 Emergency Control .;.;..,....;......,..,.., ...... cous 4 8 Salt Handling SySGems -+eeoesseen- Peeeeeaeesaaeeiaa e 4.8, 1 Introductlon,.,g};;.;;;...._ ...... Cereeanas Creereaaa, ' ;_'4.8.2 FL.el Sa,lt Prepara,tlon A 4.8.3 Coolant Salt. Preparatlon P Ceeeeneas 4.8.4 Reactor Salt Purification -e.eeviireieriieniiennnn., 4.8.5 Coolant-Salt Purification ......veseeevevevenenenenes 5. 4.8.6 viii CONTENTS (continued) Reactor Salt Charging System .... » e 0 4.8.7 Intermediate Coolant Charging System ............ e 4.8,8 UF,; Addition Facility .......c.covvvnnnnn. Ceeeseaa e 4.8.9 Fuel Salt Drain and Storage System c..cvveenne tesena 4.8.10 Coolant-Salt Drain SYStem cecvuenneenccneconcasannns 4.8.11 Spent Fuel Withdrawal System ...... Sesesaancennsens 4.8.12 High Level Radioactive Salt SAMPlEr } ‘m L\ ix CONTENTS (continued) Page 5.2.6 Cold Traps eeeeeceeocenes cereens s ecesencasaans Cerene 100 5.2.7 Reduction Reactor -.........cc... Creesssreassaacnnenns 100 5.2.8 Transfer Tanks ........ e e 100 5.2.9 Waste STOrage TANKS ««eenenonenenronsncncesacenenenss 100 5.2.10 TFreeze Valves .eoveerecrses tesessesasans Cetesseneas ... 102 5.2.11 Samplers seeeeaceeassssses Citeeeane U Cesecerrenas . 102 5.2.12 Biological Shield «..... cecessetsasnssencaan recseanns 102 5.2.13 Process Equipment Layout ..... testaesesitennacaraanan 102 5.2.14 Plant Layout +.coveevvecesn. Pheereertarnsansasesneanne . 105 5.2.15 Capital Cost Estimate ..cevvvveniennnes Cereasseraenas 105 2+2.16 Operating and Maintenance Cost Estimates ....... vee.s 113 5.2.17 MSCR Irradiated Fuel Shipping Cost ...cevvvenn.. ceee. 113 5.2.18 MSCR Unit Processing CoOSt «.eevvnn ceeens tecsssenanans 117 5.3 Thorex Central Plant ..eeveeevieennann Chreesssssaasssenarans 118 5.3.1 Head-End Treatment ...... ceersenans ceeersanns eseeseee 119 5.3.2 Solvent Extraction «c.vveseenrsersiscencsesnsacnes vaoaae 121 5.3.3 Tail-End Treatment ..... Seeeseisanas P 121 5:3.4 Processing Costs «eesse Ceeteerseraessesesteseearaasas 123 5.4 Comparison of Processing Cost Estimates ...ccceveevceneannn 125 FUEL CYCLE ANALYSIS ....... cereanans Cereecenr ettt oo 127 6.1 Analysis of Nuclear System cieeresasrasnes ceseaceennene eeo 127 6.1.1 . Computer Erogranw £y essiosrestenesesienennennse veees 127 6.1.2 Reactor Phy51cs MOGEL tevvensennaneanans Cereereeaes 129 . 6. l 3 Cross Sectlon Data ..;;,..... ..... teesccacssenencasns 129 r.6 .2 Analy51s of Thermal- and Mechanical System «eeceverianianeas 130 6. 2.1 Maxxmum Fuel Temperature ceteriiiainaenas B A X 6.2.2 Minimum Fuel Temperature «e..coceecsesnsensencsvanees 131 6.2.3 .,Veloc1ty ......{.._1g.;.,....,.....,... ..... veasesses 131 6.2i4 Fuel VOLUIE «+sevvvensnnnnesesnnnsesnsonnsessananness 133 : 6;3'_Analysis of Chemicaljsyétem‘.;....; ..... Cetesenaneneenaiens 134 6-3.1 THOTIUM-232 +evrrnrrerrnnnreesnnneeennnes erieeeeeaas 134 6.3.2 Protactinium-233 ..eeiiiiiiiiiiionans Ceeiesseaerannas 135 Page 6.3.3 Uranium-233 .vecceeerencenacens Ceavens cesesene cresne. 135 6.3.4 Uranium-234 .cececetrrecsenas cessiessssrrasssevanna es 135 6.3.5 Uranium-235 «.eeeeerecneeccenes feeiseeenases Cearaen 136 - 6+3.6 Uranium-236 ...... crecsesesssseavese veereissetsannn .. 136 6.3.7 Neptunium-237 cescceecrccscons craees ceessensas cearees 136 6.3.8 Uranium-238 ..ceeeessssessnssssscasssne taeseqsssssess 136 6.3.9 Neptunium-239 and Plutonium Isotopes ...... Cereeeaaes 136 6.3.10 Salt ........fl......;;..;.;................. ...... veo 136 6.3.11 Xenon-135 and Related Isotopes .c.ceeeceoncscnsanenss 137 6.3.12 Noble Metal Fission Products ..... ceeens cervsessseass 137 6.3.13 Other Fission Products eseeeeeccecss P K . 6.3.14 Corrosion Products cceeescccescscscsnnss ceserens coses J_3'7< 6.4 Fuel Cycle Optimization ..vceeeececicecnrnccencssnsranseases 138 6.5 Reference Design Reactor ..... teceevsanetenonnees Cecesennne 141 6.5.1 Specifications ccccteetirtieiticrcrscarenens teeane s 142 6.5.2 Neutron ECONOMY ecceccsssccacccsncns ceeseasassrseasss 143 6.5.3 Inventories and Processing Rates ....cccoevenens ceees 147 6.5.4 Fuel Cycle COSt «eeceveecscnansen Cereeseieinaaans .. 148 6.6 Parameter Studies ...... ceceeecsessersesuresatatnasans eeese 150 6.6.1 Processing Cost as Parameter .......... ceereteansanes 150 6.6.2 Effect of Xenon Removal «cecssceccsesssssnnnnns recess 152 6.6.3 Effect of Product Sale Without Recycle ..... cessecsss 153 6.7 Alternative Design and Cost Bases ..ceceverenns ceessessenes 154 6.7.1 Thorex Processing Cost Estimates «...... creseccsssees 154 6.7.2 Reactor-Integrated Fluoride Volatility Processing ... 155 6.7.3 Reactor-Integrated Precipitation Process ..eeseees.. . 156 6.8 Evolution of a Self-Sustaining MSCR --..... Ceettsereeatnana 156 6.8.1 Reduction of Leakage «oeeesesss Ceebeeraciareneenneaes 157 6.8.2 Reduction of Xenon Captures «.ceeevsverrcscsnanns eeees 157 6.8.3 Reduction of Fission Product Poisoning .............. 158 6.8.4 Improvement of Mean Eta and Reduction of 226U x CONTENTS (continued) Captures «cceeee.. cheseanes ceedtesnsiersataaseaeenanes 158 “ i3 O < !(fl \bu O o g QO > L A xi CONTENTS (continued) 6.8.5 Ultimate Breeding Potential of MSCR .eveceevencannnes 7. MSCR CAPITAIL INVESTMENT, FIXED CHARGES, AND OPERATING EXPENSE +vcccasscsonassassnassosasnsacosaasssssanssnos e sesaessessee 7.1 Introduction ........ TR 7.2 Summary of MSCR Capital Investment ..cccceceienecanacncanen 7.3 MSCR Fixed Charges «csvevneereene Cesetcsssarsesnsiaance e 7.4 MSCR Operating and Maintenance Cost Estimate ..c.coeceeneesn 7.4.1 lebor and Materials ceeessaas tecetecarsasaaanaeeanann 7.4.2 Qperatlon and Maintenance COSt «eveeseseecnesneaneens 8. RESULTS AND CONCLUSIONS .cceoeose teecesssetetasseseassananraesan 8.1 Fuel Cost «vveevesnen cesessrseraansanses cerennee cheeneasaas 8.2 Fixed Charges ...e..... eeenseeiaeans e et eeeeeeanes 8.3 Operation and Maintenance EXPENSE seevessarneeeannnnesssnes 8.4 COSt Of POWEY +eeeeecsctoesosassssnccssssosnas cereeeneas cees 8.5 Breeding Potential of the MSCR +vrerrernnnnnansasoncessaess 8.6 Conclusions ...... e et eeeeereeen e treersensacencn 8.7 Recommendations ««eeeeeseeceesasoecnses Ceenns ceseesaansanas 8.7.1 Title 1 Design Study Of MSCR teveverecccceenccncoanes 8.7.2 Conceptual Design Studies of Advanced Breeder Reactors .eeeveveens crrenasacanans cersesasessnesvrsns 8.7.3 Fundamental Studles of Alternatlve Chemical ' Processes sececescssesis cessesasseasaas ssessascsasne 8.7.4 'Englneerlng Laboratory Study of Prec1p1tat10n Processing ceveeeecncrsrsescecccctstannoesonnnnenne, 8.7.5 Pilot Plant Study of HF Dissolution Process ......... APPENDICES-.......;.;..;...4;.;...;;... ..... e teetee et in e Appendlx»A — MULTTIGROUP CROSS SECTIONS FOR MSCR CALCUIATIONS ...... Appendlx B — EFFECTIVE THORIUM RESONANCE INTEGRALS «cvovrenovsenaens - Introduction . ...,,...........,........,........;-.......,..fl Analysis .s.e.en ...,;.3.,g.;;....;...;..;...........,...,.;;....._ -Sampie Calculation ,..,.;;.3,;....;.....;... ...... esaesessaenes Results +o... et et Ceeeenaeeneranans Ceeereeranas SYmMbOLS seencrsrsernnrnans G eesesirtesesnsrasereassanes Ceerene cone References ccieseeeees e e esasanseses e e s as censs s caseseeanes 174 174 174 175 177 179 196 196 197 200 201 201 xii CONTENTS (continued) Page Appendix C — ENERGY DEPENDENCE OF ETA OF 223U ........... pevesnans . 205 SumAary -esseees ceeese tetesvsssvecsassssssesascrsssnsosssssaesras 205 References »eceevesss. Ceeeatreteeeteeereaaanas Ceeeereneas cieesss 208 Appendix D — THE MERC-1 EQUILIBRIUM REACTOR CODE +vvvee-n.... Meeees 209 Introduction .ececevene cessnans et ieraeaeas crseesee 209 TheOrY ceccsececssesensscnsonsanssncanas ceeseecscns cresnacss e 209 References «eeeeeesss et eeeetniaraeearieeneas tessestcennanns 220 Appendix E — FISSION PRODUCT NUCLEAR DATA «.ccoen.n. Ceeeeees e 221 , References .............{...,.;..;..........................;...' 224 Appendix F — TREATMENT OF DEIAYED NEUTRONS ..... Cereeeen Ceeeeeans . 225 SUMMATY «sesovsoensasseasassssencsncsncsncanssassssessnrsansanses 225 References «ceceeceas teecscsececsesccscstrsase coesecnssese sesesss 229 Appendix G — TREATMENT OF XENON ABSORPTION IN GRAPHITE ............ 230 Introduction ...... e e e, .. 230 Analysis ceeceeevscccncnnnss Cesesessans P Ceecean crsessenasesases 231 Reference e..cseesees cecsacnsena ceccscsetssesnascssesrascansaces 234 Appendix H — THE EQUILIBRIUM STATE AS A BASIS FOR ECONOMIC ' EVAHIA-TION OFTHORI[JMREACTORS 4 & & 5 6 & 580800 LI B I O B B N I B N I BN BN RN N 235 IntrodUCtion teeceecssessesscscnsansessassscssscsssccsssssa ceececeasne 235 Methods ...... crsscsresnnes sesesesacnstennenas ceectssrasnas ceres 236 l. Fission Products ececeses cecsesemenennanesnans cesccccns eve R36 2 3. Uranium=234 cecesreessassscsasossasessccnsss ceeerecassenss 239 e Uranium-235 cececesessesasssessssscsasssanans creteeneness 239 5. Uranium-236 «eeeeesecrceccocnns Cereeeees Ceeetrenraeneesss 240 6. Uranium-238 ........ tetetesneasscnesnannrennns ceeesasaens 241 Results andjbonclusions hessascaseascrsestsrarensatssranns ceress RA42 1. Fission Products «ceeeceeesecaccassacscnnane cesrienas cenee 242 2. Uranium-233 ..... cescesessenssssssseasenans tesesanctasenns 249 3. Uranium-234 ceeeves . cecesecrescans 249 4. Uranium-235 ....... Cettesaasasiseneasessassaseananna ceees 249 5. Uraniume-236 ccceceevessrsessssssssssssassssssssssoss ceseans 249 6. Uranium-238 teeiiiiertsssssacrsssassccssssansnnnnnss cevees 249 v Urniume233 «eceeereveennens e eeerenneenn X 14 D vy I O d ;‘5 1) xiii CONTENTS (continued) Q%‘ y o | { Page Appendix I — ESTIMATES OF PHYSICAL PROPERTTES OF LITHIUM- BERYLLIUM MSCR FUEL AND COOLANT SALTS tvieivecrrorasnocsassnasonnons 252 Introduetion cereererrioessassnsssasccssncesoncnnenns cevessennsea 252 Viscosity «.eees cesaes tesessstesaetarecaavens cearessnnn crieessess 253 Heat CapacCiby ceveeveveeecensrennsrsncressassasssasaannse crsessra 256 Thermal Conductivity .cieeececenns. beeresinane cecenes ceeereanans 256 Density ceeeesececeas Cetasseasseeioteanancnne ceereae ceeseerassan 258 Liquidus Temperature «veceessesscesacsanaans Cereesiassannesennn . 258 = References .svoeeeesseens s4esuesrensssrsessstasssnenna cesescnaeses 260 < Appendlx J — FUEL AND CARRIER SALT COST BASES .evevevse cesaaas ceees 261 TIET OAUCELON + v e e s eveennenennenennsesenensenensenansenensennns 261 Bases for Establishing Prices ....cvavnns sesseses tesseteanaeaans 269 General Comments on Price Quotations ............ Perereeetaannas 269 Thorium Fluoride ....cceuenn. creens ceenna eseesrcssasesewrennns 269 Zirconium Fluoride sceeveecevesns crecsese teerseccassssnecnsssas 269 ‘Beryllium Fluoride ccoeseceescecsnecanonens Meesresesrasean e 270 Sodium Fluoride ..... teecsccesssessesssssssssassansans ceeenns 270 Thorium Oxide ........ ceesrassesa cecesseresna cieeaneann veeaee 270 Lithium Fluoride teceeeeeescesaccnnses Ceerecrsessearatsanas .. 270 Recommended Values for Molten Salt Fuel ......... ceeeseavan ceees 271 i References +secesececas tecesenorace N cresersns daacans 272 T Appendix K — MSCR POWER LIMITATION RESULTING FROM MODERATCR ] THERMAL, STRESS +cveeetvescsssnsessarsascssorssnennsns tessressasnanea 273 B Summary ...............Q;.,.ff....;...;. ..... R ceneeas ceeee. 273 , Réferefices ......;Q;;...;.;...Q.;..........;.......,............ 279 - Appendlx L — VOLUMES OF FUEL SALT AND INTERMEDIATE COOLANT SALT FOR 1000 Mwe MOLTEN SALT CONVERTER REACTOR ceresessrssaanaasiaeseass 280 Introductlon ......................................;..}......... 280 | NEl Salt VOl'U.me -e;go--..-..oonoc--o-oooo-..a-fo-o_oiood-ooo;'--t 280 o Coolant S alt vo lume o " P e & ' .99 0 0 & 0 O O 20 S8 e RS BB S e E e .' _. * ' ..... 282 - Appendix M — EVALUATION OF A GRAPHITE REFLECTOR FOR THE MOLTEN . } SALT CON.VE:R.Tm RMCTOR ..... & & 9 % & PSP B e P e ® & & 5 8 & & 4 O 50 & BB P Y S e hAe 285 g” THtTOAUCELION = vosvnnnnns e e s 285 AW:;RW xiv CONTENTS (continued) . | . Page Results ...'C.......'.._VCOD'OQIO..l..l..l.lCQ:l..ll.....‘......g". 286 ConCluSionS ..I.'.‘....I..;....'..'...I‘...;.l'.....!l...'.i.." 288 Appendix N — DETAILED ESTIMATE OF 1000 Mwe MSCR CAPITAL ‘ mESMNT & ¢ 8 8 B 2% 8 S E B0 d P ¢SRS SE S e e su........lloooolii 289 SUIMMATY «+ e cepeennernesionsonsssssonanseosasnncaacsssassssscnces 289 Investment Reqnifements Cetereeerecieiananan cesessasnessssvssses 290 Appendix O — DESIGN REQUIREMENTS FOR THE MSCR MODERATOR +«evereasse 328 INtrodUCtion tieceescrssonerensasurosssvosansncesssssncascosasnes 328 Moderator .,_,_“,_,,.:,_,,,_,,,“_,,,,,,,__,_.“.,,,,,_,_,,,_,,,”,,' 330 VOid Fraction -..-.o‘-oo-ooscoouoo‘-o-o-ooooo'iooto--onoo-o--oono-n. 330 - Permeation of Graphite by Salt .......cccciivniiiiiiiiiioneeaess 331 Graphite Shrinkage ccccccc N T E TR RN .po.o--o-cooo--o-i ....... 332 Graphite Replacement .cecevscecccsccerscssssecscscsasssnsasscsssee 333 DifferentialExpanSion ll...'.....'l.'..q.....ll.r.itl...‘.'.l..! 335 References ¢ 00 OB EE LSS SRR E S DS SSE S SNTES SN E S S SO SENA eSS eSS 336 BIBLIOGRAPI{Y © 5 5 062 289 T 0 S0 SEEEE SN NE NSRS ENSNeSLSEEBOIBEBSES 337 O ( " iy 1) £ s)‘_- » a AP v » XV ACKNOWLEDGMENTS The comments, suggestions, and critical reviews of R. B. Briggs, P. R. Kasten, J. A. Lane, H. G. MacPherson, E. S. Bettis, A. M. Perry, and S. E. Beall are gratefully acknowledged. The computer programmihg was performed by J. Lucius, ORGDP Computing Center, and many of the drawings were prepared by H. MacColl. The capital costs were estimated by C. A. Hatstat of Sargent and Lundy, Engineers. R. P. Milford and W. G. Stockdale, ORNL Chemical Technology Division, assisted the chemi- cal processing cost studies. Roy Robertson and I. Spiewak, ORNL Reactor Division, assisted the design and analysis of the energy conversion systems. Special studies reported in the Appendix were performed by R. H. Chapman, J. W. Miller, and C. W. Nestor of the ORNL Reactor Di- vision, and D. B. Janney of ORGDP. W > » L) MOLTEN SALT CONVERTER REACTOR - Design Study and Power Cost Estimates for a 1000 Mwe Station % L. G. Alexander, W. L. Carter, C. W. Craven, D. B. Janney, [ T..W. Kerlin, and R. Van Winkle ABSTRACT The MSCR is a one-region, one-fluid, graphite-moderated converter reactor fueled with a mixture of the fluorides of thorium uranium, lithium-7, and beryllium which is circulated through the 20-ft-diam core to an external heat exchanger. Heat is transferred through an intermediate salt-coolant to steam at 2400 psi, 1000°F in a Loeffler boiler system having a net thermal efficiency of 41.5%. Spent fuel is processed by fluorination (at 0.08 mill/kwhe) for recycle of isotopes of uranium. The stripped salt is discarded. ,'i w) A capital investment of $143/kwe (3.0 mills/kwhe), an operation and maintenance annual expense of $2.1 million (0.3 mill/kwhe), and a minimum fuel cycle cost of 0.7 mill/ kwhe (optimum conversion ratio is ~0.9) were estimated, giv- ing a net power cost of 4.0 mills/kwhe. All costs were based on 1962 bases ground rules. | Second generation plants may have capital costs as low as $l25/kwe. Conversion ratios slightly greater than one can be obtained in advenced designs. | This study was completed in December 1962 and does not - - reflect increased feasibility and superior performance of ' two-region, two-fluid molten salt breeder reactors made pos- sible by recent (January 1965) advances in core design, heat transfer, and fuel-salt processing. V'iilaf',SIHflNM&fY - The Molten Salt Cbnvefter'Reacgor_(MSCR) is a one-region, one-fluid, near-term reactor that does not require any technology beyond the scale-up - of that already developed at ORNL or to be demonstrated in the MSRE.- Sa- lient characteristics are given in Table 1.1. ) -~ Table 1.1, Characteristics-of,the Molten Salt Con#erter Reactor Thermal capability Net thermal efficiency Diameter and height of core Moderator | Volume fraction of fuel in core Composition of fuel carrier salt (mole-percentages) Density of fuel salt Heat capacity of fuel salt Velocity of fuel salt Inlet'temperature OQutlet temperature Flow rate Volume of circulating stream Power density in core (av) Power density in fuel salt (av) Thorium specific power Fissile material specific power Fertile material exposure Intermediate coolant (mole-per- centages) Steam conditions C: Th atom ratio Th: U atom ratio Mean neutron productions (fj€) Optimum conversion ratio 2500 Mw 41.56 20 x 20 ft Graphite 0.10 68-LiF, 22-BeFp, 9-ThF,, 1-UF, 190 1b/ft3 0.35 Btu/1b*°F . 6 fps 1100°F . 1300°F | 160 £t>/sec © 2500: £t3 14 w/cm? 35 w/cm? 30 Mwt/tonne 0.9 Mwt/kg (47 Mw days/kg 63-LiF, 37-BeF 2400 psi, 1000°F ~300 ~30 2.21 0.9 O » (i‘ ) .g'l ) R} ) 1.1 Description The reactor vessel is fabricated of INCR-8 alloy and is filled with cylindrical graphite logs 8 inches in diameter and 24 inches long. The fuel, a mixture of the fluorides ef 7Li, Be, Th, and U flows upward through the pasSages around the 1bgs and is discharged through eight pumps to an equal number of heat exchangers where the heat is transferred to an inter- mediate-salt coolant. Saturated steam is superheated in a shell-and-tube exchanger; part of the steam is routed to the turbines; the rest is re- circulated to Loeffler boilers where saturated steam is generated by in- Jecting the superheated steam into water. Thus, thermal contact of the coolant salt with subcooled, boiling water is.avoided, and_thermelfstress in the tube walls is tolerable. The thermal efficiency is in excess of 40%. Twenty-five hundred Mw of heat are extracted from a single core at average power densities in the fuel salt of not more than 35 w/cmB- 1.2 Fuel Reprocessing Irradiated fuel is‘removed from the reactor daily, collected into processing batches, and treated with fluorine for recovery of isotopes of uranium (fully decontaminated) as the hexafluoride. The stripped salt is discarded. Recovered UFg is reduced terUF4, blended with fresh salt, and recycled to the reactor. Net burnup and loss of fissile materiél-are com- pensated by addition of 95% enriched 235y. 1,3--Nuciearfend~Thermal Performance The limiting crlterla (e g ; max1mum ellowable fuel temperature, maxi- - mum allowable thermal stress in graphlte, etc ) were chosen’ conservatlvely “throughout, and pr0v1de cons1derable margin for improvement in later de- - signs. ) The key variables (core dlameter, volume fraction of fuel in core, ,carbon,thorlum ratio, and proces51ng rate) were optimized w1th respect to the fuel cycle cost. Characteristics of the optimized system are listed’ in Table 1.1 where it is seen that the optimum conversion ratio is 0.9, with slightly permeable graphite that absorbs 135Xe only slowly. 1.4 Fuel Cycle Cost The estimation of inventory and replacement charges for the MSCR is straightforward. Processing costs are less well defined; however, the processing contributes only a small part of the total fuel cost, and the aggregaté is not sensitive to large errors in the processing cost esti- mates. A central Fluoride Volatility facility capable of processing 30 ft3/ day of salt was designed and costed. Only isotopes of uranium are re- covered; carrier salt and thorium are discarded along with fission pro- ducts. Unit costs and the components of the fuel cycle cost are listed ~in Table 1.2. Table 1.2. PFuel Cycle Cost in 1000 Mwe Molten Salt Converter Cost Bases Capital investment in processing plant: Reactor Plant Annual operating expense: $2 million Turn-around-time: Batch s Shipping costs: 2 days ize: 6000 kg Unit processing cost: $27/kg Th $10/kg Th Purchase price ThF;: $19/kg Th Carrier salt purchase price: $1130/ft3 Fissile isotopes: $12/gram $26 million Charges, mills/kwhre Material — Total Inventory Replacement Processing TH232 0.033 0.043 Pa2?3 0.008 U233 0.183 0.082 y?35 - 0.037 0.156 _ —_ . Total 0.262 0.199 0.082 0.54 Salt 0.062 0.079 0.14 Total charges, mills/kWhre 0.7 {» Py € ) LY i‘b 1.5 Power Costs The cost of power was obtained by combining fuel-cycle costs with estimates of capital charges prepared by Sargent and Lundy, Engineers (95,96), from a design study conducted at ORNL. Equipment was sized and specified in sufficient detail that costs might be estimated by usual proceduresé Plant arrangement drawings were prepared from which costs of buildings, piping, services, etc. were estimated. Operation and mainte- nance costs were estimated according to standard procedures (52). A sum- mayy of the principal items is given in Table 1.3. Table 1.3. 1000-Mwe Molten Sait Converter Reactor Construction Costs Direct construction costs Structures and improvements $ 5,997,950 Reactor plant equipment - 51,324,350 Turbine-generator units 26,843,700 Accessory electric equipment 4,375,300 Miscellaneous power plant equipment 799,900 Total direct construction costs 89,341,200 Indirect costs 9,083,300 Engineering design and inspection costs 15,080,300 Miscellaneous charges 35,370,800 ~ GRAND TOTAL $148,875,600 Net station power . , | 1038 Mwe ‘Unit cepital cost =~ o | $143/Kwe The fixed Charges-(14-46%):onithe[capital investment contribute 3.0 mills/kwhre to the power cost. The uncertainty in this cost might run as high as 15-20%, and the fixed cherges might renge up to 3.5 mills/kwhre. ~Operation and maintenance contribute 0.3 mills/kwhre to the total power cost (Table 1.4). Because of the many uncertainties, this estimate mey be low, and the cost might run as high as 0.5 mills/kwhre. 6 Table 1l.4. 1000-Mwe Molten Salt Converter Reactor Operating and Maintenance Cost Wages and salaries $ 872,000 Routine materials 220,000 Maintenance 800,000 Management _262,000 Total $2,1.54,000 The various contributions to the cost of power have been summed in Table 1.5. Table 1.5. Cost of Power in a 1000-Mwe . Molten Salt Converter Reactor Charge, Ttem mills/kwhre Fuel cycle cost 0.7 Fixed charges 3.0 Operation and maintenance 0.3 Cost of power, mills/kwhre 4.0 Taking the upper bound on these three items estimated above (fuel cost ~1.0, fixed charges ~3.5, operation and maintenance ~0.5) gives an upper limit on the cost of power of 5.0 mills/kwhre. 1.6 Advanced MSCR The system evaluated above was based on the scale-up of current tech- nology, and was conservatively designed in every respect. There are sev- eral obvious improvements that could be incorporated into a "second gen-. eration" design. If the design criteria were relaxed, metallic sodium could be substituted for the intermediate salt coolant (saving about $10 million in capital costs. This would also permit the use of "conventional O * "’ o) iy e e I e L A AL 51 8 R AP B 1t . bt . *J) 7 once-through sodium-heated boilers and reduce the cost of the energy con- version system by about another $10 million. The total cost would then be ~$125/kwe. By careful design and development the fuel volume might be re- duced from 2500 ft3 to lSOO{fté. Separated °?Mo could be used to clad the graphite and so reduce absorption okaenon therein and also as a struc- tural material by means of which a blanket of ThF, bearing salt could be added at the periphery of the core to reduce neutron leakagé. The use of Fluoride Volatility coupled with the HF Solution Process to remove rare earths could reduce the fission product poisoning to very low levels while permitting recycle of carrier salt (but not thorium).. Pfeliminary calcu- lations show that these improvements (all within reach of modest develop- ment programs) might increase the conversion: ratio above 1.0, and, with ~ the reduction in capital costs noted, result in a power cost of 3.4 mills/ ‘kwhre. 1.7 Post Script — January 1965 This study was completed in Décember-l962, and does not reflect in- creased feasibility and superior performance of fiwo-region,'two-fluid molten salt breeders made possible by the recent advances (January_l965) in core design, heat transfer, and fuel-salt processing alluded to in the Foreword. 8 2. INTRODUCTION 2.1 Purpose, Scope, and Method of Approach - The purpose of this study was to evaluate the economic potential of a near-term molten salt power reactor. "Near-term" characterizes a system which utilizes only techniques or equipment currently under development. 2.1.1 Figure of Merit The economic potential of power reactors is measured by‘the'net cost v of electric power. | Fuel cycle cost, although not definitive, is also an important index of economic potential. Moreover, the optimization of the fuel cycle is a required first step in the detailed design of both reactor and electric ~plants. In this study, the reactor and its associated heat transfer sys- tem, the energy conversion system, and the fuel reprocessing plant were designed in detail sufficient to permit the optimization of the fuel cycle. ; 2.1.2 Reactor Concept A concept was selected for evaluation, which, judging from previous experience, would satisfy the "near-term" requirement and yet would ex- hibit attractive fuel costs: A single-fluid, single-region, graphite- moderated molten-salt reactor generically related to the Molten Salt Re- actor Experiment. Since the breeding ratio was expected to be'less than A unity, the system was designated the "Molten Salt Converter Reactor" = «wu- (MSCR). ' | - 2.1.3 Procedure In & series of preliminary calculations, therlimitations on reactor design imposed by consideration of allowable témperature, pressures, ve- locities, thermal stress, etc., were determined. DeSign and cost bases were established, and the fuel cycle cost was minimized by optimization of the key variables, which in the MSCR are the core diameter, carbon/ thorium ratio, volume fraction of fuel in the core, and spent fuel \fiJ o 4 ) -« [B) pg» 19 processing rate. For the optimum conditions, the fuel cycle costs result- ing from alternate bases and assumptions (e.g., removal of xenon) were determined. Finally, the ultimate performance resulting from & concatena- tion of all favorable assumptions and potentially low processing costs was estimated. 2.2 -Status of Molten Salt Reactor Development 2.2.1 Early Work Molten salt fuels were conceived originally as a means of satisfying the requirements for very high temperature and extremely high power density necessary for aircraft propulsion. A very large amount of work on the physical, chemical, and engineering characteristics of wranium and thorium bearing molten fluorides was carried out as part of the ANP program at Qak Ridge National Laboratory The technology of molten salt reactors was first 1ntroduced 1nto the open literature in 1957 by Briant and Welnberg (14). Papers by Bettis et al. (6,7) and Ergen et al. (31) reported the Aircraft Reactor Experi- ment, a beryllium-moderated reactor fueled with UF, dissolved in a mix- ture of the fluorides of sodium and zirconium, and contained in Inconel. The reactor was successfully operated in 1954 for about 90,000 kwhr with- out incident at powers up to 2.5 Mwt and temperatures as high as l650°F, The potential usefulness of molten salt fuels for civilian power was recognized from the start.. The features that attracted attention were the high temperature of the fuel (permlttlng use of modern steam technology and attainment of high thermal efflciency) combined with a lOW‘V&POr pres- sure, the hlgh stablllty of hallde salts under radlatlon, and the a&van- ~ tages that a fluid fuel prov1des._ These include a negatlve temperature '"coefflclent of reactrv1ty, absence ‘of the need for 1n1tial ‘€XCess reac- ' ~tiv1ty and of neutron wastage in control elements, no limitation to fuel exposure due to radlatlon damage or fuel burnup, the absence of & compli- cated structure in the reactor core, removal of the heat transfer opera.- tlon from the core to an external heat exchanger, and the potentlal for a low-cost fuel cycle. In addition, suitable molten salt mixtures exhibit a 10 solubility for‘thorium fiuoride sufficient for all reactdr'applications; moreover, these mixtures may be economically and rapidly processed for the recovery of 233U by means of the well-develoPed Fluoride Volatlllty Process. ' Studies of‘powei reactors utilizing molten salts have been reported by Wehmeyer (109), Jarvis (49), Davies (27), and Bulmer (15). Davidson ~and Robb (26) conceived many of the features of one-region thorium con- | verter reactors and anticipated some of the development prbblems. 2.2.2 The Molten Salt Reactor Program The molten salt reactor program. was 1naugurated at ORNL in 1956 (57, 58) to exploit the technology of molten selt fuels for purposes of economic civilian power. Several parts of the program were: (a) a reactor evalua- tion study to select the most promising concepts for c1v1lian power and to plnp01nt specific development problems; (b) an extensive materials de- velopment program for fuels, containers, and moderators; (c) an equally extensive program for the development of components, especially'pumps, valves; and flanges suitable for extended use with molten salts at 1300°F; (d) a modest program for the discovery of supplementary chemical processes for recovering valuable components (other than uranium) from spent fuel; (e) a program for the development afid definitive demonstration of the feasibility of edmpletely remote fiaiqtenance'of molten salt reactor sys- tems; and presently (f) the Molten Salt Reactor Experiment (MSRE). 2.2.3 Fuel Development The program for the development of molten salt fuels in the Reaetor Chemistry Division at ORNL has been highly successful (56). The five-com- ponent mixtures (fluorldes of Li, Be, Th, U, and Zr) developed for the MSRE (12) have many exceptional features. They have melting p01nts well below 1000°F, with ample solubility for UF,, ThF4, and fission product '.fluorides. They are thermodynamically stable with vapor pressures less than 0.1 atm at temperatures well above 2000°F, and, being ionic liquids, | are not subject to permanent radiation demage (e.g., rediolytic dissocia- tion) when in the liquid state. The parasitic capture Cross sectiensfof the base elements (71i, %Be, and 1°F) are satisfactorily low, and "Li O ¥ § ” O 5 +) {4 iv #) 11 is available at attractive prices in grades containing as little as 0.005% Li-6. The high volumetric‘heat capacities of salt mixtures make them better heat transfer medis than most liquid metals in spite of the higher film conductances obtainable with the latter. ' These mixtures do not appreciably attack the container material (INOR-8), corrosion rates being less than 1 mil/year (possibly as low as 1/2 mil/year) at temperatures below 1300°F (28). Although it is not now anticipated that it will be necessary to use INOR in the neutron active zone, since the moderator material (graphite) is suitably self-supporting, experiments have shown that the corrosion is not appreciably accelerated by radiation. A long life (10-30 years) is predicted for all components constructed of INOR (reactorrvessel, pumps, heat exchangers, etc.) be- cause resistance to corrosion does not depend on maintenance of a protec- tive film but stems from the inertness of the base metal toward the salt. Molten salt fuel mixtures are compatible with graphite. Tests of a typical grade show that the salt does not wet the grephite and penetra- tion is mostly confined to the surface layers (84, p. 93). Some CF,; has been observed in post-irradiation examination of in-pile experiments. Since CF, is thermodynamically unstable with respect to the salt, it is thought that its formation resulted from attack on graphite by free fluo- rine produced by radiolysis of solid salt. Since the fuel-salt must be maintained in the liquid state for other reasons, free fluorine would not normally be present in the circulating stream. Xenon is not adsorbed'appreciably'on graphite (17) at reactor tempera- tures, though it will saturate the #6ids present because of its extremely Jow solubility in salt (107). However, it may be,pcssible to exclude Xe from the graphite by treating the surface to close the pores there and render interior'poresfinaccessibie (5). Purging the'salt_with a stream ,¢f,héliflm in the pump bowl or in .a special contactor would then maintain . the Xe concentration at a very low level (Section 6.8). TIodine remains in the ionic state and is not absorbed. -Noble metal flss10n products are expected to be reduced by INQR outsmde the core. o . The phase behavior of a great many mixtures has been 1nvest1gated (108). Proposed mixtures containing up to 40 mole % BeF, have viscosities 12 | adequately low and dissolve heavy metal fluorides (UF4, ThF,, or ZrF,) in concentrations up to 15 mole % with liquidus temperatures less than 1000°F (56). Additions of 5/ mole % of ZrF, to the base salt satisfactorily re- duces the sensitivity of the fuel mixture toward precipitation of UO; by oxygenated contaminants (e.g., air, water, lubricating oils) which will be difficult to exclude entirely from a large reactor system. Graphite is readily de-oxygenated by in situ decomposition of NHgF-HF vapor, which shows negligible attack on the INOR. o — _ Thermophysical properties of the important salt mixtures have been measured (8,24) in detail sufficient to permit reliable calculation of pumping and heat transfer characteristics, which are good. _No evidence of the deposition of scale or dendrites in the heat exchangers has been “found. 2.2.4 Container Development The development of nickel-molybdenum base alloys (INCR series) for containment of molten fluorides was conducted jointly by ORNL and Inter- national Nickel Compeny. In addition to the resistance to corrosion men- tioned above, the alloys have good—to-éxcellent mechanical and thermal . characteristics, (superior to those of many austenitic stainless steels) ~and are virtually unaffected by long-term exposuyre to salts or to air at 1300°F (12). The alloy has been made'by several major manufactUring com- panies, and it is presently available on & limited commercial basis in the form of tubing, plates, bars, forgings, and castings. Exhaustive tests at ORNL have shown that its tensile properties, ductility, creep strength, cyclic fatigue strength (both thermal and mechanical) are adequate for molten salt reactor applications when judged in accordance with criteris used in the ASME Boiler Code (75-87). INCR is weldable by'conventional techniques using welding rods of the same compositidn as the base metal. A gold-nickel alloy has been developed at ORNL suitable for remote brazing of reactor components. INOR begins to soften above 2000°F and melts at 2500°F. The thermel conductivity is about 12 Btu/hr-ft:°F at 1200°F. No major difficulties have been encountered in the design and fabrication of reactor components, including pumps and heat exchangers (12). T o E2) L] £ 13 2.2.5 Moderator Development ‘Graphite, because of its good moderating properties, low neutron cap- ture cross sectiOn, compatibility,with fluoride salts and INOR, and excel- lent high-temperature physical properties is a superior moderator for molten salt reactors. The graphite proposed for use in the MSRE has a density of 1.8 g/cc and a kerosene-accessible porosity of 6%. About half the pore volume is accessible from the surface. However, as mentioned above, molten fluorides do not wet graphite and permeation of MSRE grade graphite by the salt is less than 0.5% by volume at 150 psi (84, p. 93). The coefficient of permeability by helium at 30°C is 10-° cm?/sec, and Xe will be adsorbed rapidly. _However,_techniques for reducing permeability are being developed. Samples of high-density graphite having permeabilis. ties at least two orders of magnitude lower have been made (107). Development of graphites and graphite bodies is being carried out cooperatively with National Carbon Company. Pieces of graphite are pres- ently avallable in sections up to 20 in. square and 20 ft long. Graphite having outstanding mechehioal properties is available in the form of readily machinable rods, tubes, slabs, and spheres. The effects of nu-" . clear rediations4on this material are not fully known. The thermal con- ductivity declines, but probably not below 15 Btu/hr:ft:°F. Thermal stress considerations thus affect the design of moderator elements; the allowable stress is thought to be at least 2000 psi and the allowable strain at least O. 1%. These~1imits appear to be oompatible with the thermel and nuclear requlrements of optimum core design. However, experi- - mental verlflcatlon of- these ‘values is needed. At the temperatures encountered in molten salt reactors, graphite 7 w111 shrink during exposure to fast neutrons. Where: large: gradlents in ~the fast‘neutron flux exlst -the. resultlng differential shrinkage will result in deformatlons, or, 1f these are restralned in stresses. The --problem,of designlng a long-llved core structure of large pieces of greph- ite is presently unresolved.'rThe boWing'of graphite. stringers might be 'L_restralnefi by use of molybdenum hoops, but thlS solutlon mey not be suit- able for large power reactors. 14 2.2.6 Component Development Development of components for molten salt reactors has been in pro- gress for over ten years. The most nbtable-achievements to date are.the demonstration of the long-term reliability of pumps operating at 1300°F, - including pumps having molten-salt-lubricated bearings, and the demonstra- tion of the reliability and maintainébility of remotely operated freeze flanges and freeze valves. 2.2.7 Reactor Vessel No difficulties were encountered in the design or.fabricatibn of the reactor vessel for the MSRE. In large power reactors provisiOn to limit thermal stress by means of therma;\shields may be necessary, but mechanical stresses are not important because pressures greater than 200 psi are not encountered anywhere in the systems. Corrosion does not appear to be a problem. 2.2.8 Molten Salt Pumps Molten salt pumps have been operated continuously for 33 months at temperatures above 1200°F. A sump-type pump having one salt-lubricated journal bearing has logged more than 12,000 hours of operation. at 1225°F, 1200 rpm, and 75 gpm. After it was stopped and restarted 82 times, examis - nétion of the bearings disclosed no discernible attack. The use of salt- lubricated bearings will enable the shaft to be lengthened so that shield- ing may be interposed between the pump bowl and the motor with its oil- lubricated bearings. The impellers of these pumps also withstand attack indefinitely under operatifig‘conditions. It is believed that pumps of the types developed can be made in large sizes for use in large molten ‘salt reactor plants and that these can operaté'at the temperatures re- quired. 2.2.9 Molten Salt Heat Exchangers and Steam Boilers : ’ The design and fabrication of exchangers for transferring heat from fuel salt to an intermediate coolant salt are straightforward. Heat ® N i 5 i 3 w) - 4} 15 transfer experiments conducted at ORNL with unirradiated salt verify the correlations used to predict the performance. Scale did not form on the heat transfer surfaces. The Loeffler boller seems especially suited for use with molten salts. Here dry saturated steam is superheated in alsalt-to-steam exchanger; part of the superheated steam is routed to the turbines, and part is recircu- lated through an evaporator producing saturated steam for recycle to the exchanger. - Problems in boiling burnout, thermal stress in the exchanger tubes, and freezing of the salt are thus avoided. However, a fuel-salt boiler presently in the conceptual stage has many potential advantages. In this concept, the fuel downcomer annuius inside the reactor vessel is widened t¢ accommodate several hundred INOR thimbles. Bayonet.tubes,.into which water is introduced, are inserted into the thimbles, but are separated from the thimble walls by a narrow annulus filled with an inert salt. Calculations show that the heat trans- fer is adequate to produce steam at 1000°F and 2000 psi. Yet the salt and steam systems are isolated from direct contact and the salt system is . under negligible pressufe- Should either system leak, this would be de- tected immediately by monitors in the inert salt system.- Such a boiler has many advantages, including the complete elimination of one cooling loop and its associated pumps, heat exchanger, etc. 1In addition, the fuel circuit is appreciably shortened in comparison to a "spread-out" system. The steam produced will be considerably less radio- active than that prodaced in“a direct cycle boiling?water reactor. 2. 2 10 Freeze Valves and Freeze Flanges Although the hlgh meltlng p01nt of a molten salt regctor fuel (800— 71000°F) is. a disadventage in that the system must be preheated before t,fllllng and provision must be;made to avoid freezing, there are also bene- | 'fits'that.accrue; Among these ie'the fact that if a leak does occur there C T is little tendency for the materlal to disperse. rapldly :Ndble gas fis- ';_Slon products do not accumulate in the llquld and the fluorides of the "'remalnlng fission products have negllglble vapor pressure and are retained. 16 The ready solidification of salts has also been put to use in the development of flanges and valves. The remote manipulation of reliable- freeze flanges has been successfully demonstrated in many tests and in a remote maintenance development facility. Freeze valves have no moving parts, no seals, and have been demonstrated to be satisfactory inisalt = transfer and drain: pipes. 2.2.11 Molten Salt Instrumentation and Special Equipment Conventional equipmenfi is adequate for measuring the nuclear behavior of molten salt reactors; however, special equipment for handling molten salts was developed at ORNL for the MSRE. For measuring liquid level in the pump bowls, for example, a ball-float suspending an iron bob whose position is sensed by an external induction coil was developed. A single ‘point electrical probe device has also been developed for use in the fill- and-drain tanks to calibrate the weighing system. A sampler-enricher device is being tested whereby fresh fuel may be added to the fuel stream during operation, and a sample of spent fuel may be removed without contamination of the fuel stream by air or water vapor and without the uncontrolled escape of any radioactive material from the reactor. Clam-shell electrical pipe heaters for lines carrying molten salt have been developed. 2.2.12 Remote Maintenance Because of fission-product contamination and induced activity in components and piping, the fuel-containing portions of molten salt re- actors cannot be approached for direct maintenance even after draining and flushing. Semi-direct maintenance through a shield plug with long- handled tools is possible for some items, but it is necessary to develop completely remote tools and methods for many of the larger components. These include tools, techniques, and procedures for removing and replacing all major reactor components, including the heat exchanger, primary fuel- pump and motor, reactor vessel, and fill-and-drain :tank. Such equipment and techniques successfully demonstrated in the Molten Salt Remote Main- tenance Development Facility at ORNL (65). This facility simulated a F4) *) 17 20-Mwt molten salt reactor system and comprised a mockup of the reactor vessel, a mockup of the héat exchanger, together with full-scale pumps; flanges, valves, electrical heaters, thermocouples, etc. All maintenance operations were performed by a single operator from a remotely located control center, using closed-circuit stereo-television for viewing. The manipulator was a general purpose, medium duty, electro-mechanical "arm" which performed a variety of functions easily and efficiently. It was used to connect and disconnect'tube and electrical connections, to carry loads weighing up to 750 1lbs and to manipulate tools. Eight basic mo- tions, five for the arm and three for the crane bridge, were controlled independently by two pistol-grip handles on the control console. Two types of remotely interchangeabie grasping devices permitted a variety of objects to be handled. | Tools developed for remote manipulation included impact wrenches, a torque tool and bolt runner, écrew jacks on the heat exchanger for working the freeze flanges, and miscellaneous devices such as lifting slings, socket extensions, hooks, fingers, etc. All these were operated by the manipulator. In addition, a reactor-lifting jig, a pump~lifting eye, and socket extensions for the torque tool and bolt runner were positioned by the manipulator, but operated by the crane or by their own power. The installation of microphones at strategic locations inside the reactor cell to enable the operator to listen to pneumatic and electric motor sounds was found to be helpful. ) Reliable, quick1y_actingfdiséonnects for electric, pneumatic, oil, and other services were adapted dr'developed. 7 ' The cbmponents of the Remote Maintenance Facility were removed and ‘replaced several times'befofehthe system was filled with salt in order to - defielop procédures and test the'tools.. Finally, the system'wasrfilled with.salt, brought tO:tempetature_with salt circulating freely, then shut ‘down and drained. All eqfiipfiefifinfias then removed and replaced remotely, 'and'tested. The salt was'iepiééed:and brought to temperature agaih;. Ttems | "maintained?.in this way inclfidéd:the pump motor, the fuel pump, the re- actor Véssel, the heat'eXChangér;-the'filliand—drain tank, electrical pipe heaters, and'thermocouples. The . demonstration was entirely successful. 8 - Maintenance of the MSRE will be accomplished by means of the tech- niques and tools developed and supplemented with some semi-direct main- tenance operations through a portéble shield having a rotatable pilug. Long-handled tools may be inserted through this plug and manipulated by hand. These means of maintenance will be thoroughly tested in a full- scale mockup of the MSRE now being constructed at ORNL. | 2.2.13 Chemical Processing of Molten Salt Fuels The use of fluid fuels in nuclear reactors provides an opportunity for continuously removing fission products and replacing fissile isotopes at power. Thus, it is possible to hold fission-product neutron losses to th low levels and to eliminate capture of neutrons in control rods. The "Fluoride Volatility Process" is in an advanced stage of develop- ment; a pilot plant for general application is now in operation at ORNL. Other processes are being sought, and prospects are good that simple and economic means can be found to separate fission products continuously from spent fuel salt. 2.2.14 Fluoride Volatility and HF Solution Processes While the fluoride volatility process was not developed specifically for use with molten salt fuels, it has been verified in laboratory experi- ments conducted at ORNL that it is applicable for removal of uranium from fluoride mixtures containing ThF; (16). In this process, elemental fluo- rine, diluted with an inert gas, is bubbled through the salt. UFs; is converted to UFg which is volatile at the temperature of operation (500~ 700°C) and passes out of the contactor to be absorbed reversibly in a bed T of sodium fluoride. The off-gas is cooled, stripped of ngin a-scrfibber, and passed through'charcoal beds where fission product gases are absorbed. The fluorides of a few of the fission products. are also volatile but # these are irreversibly absorbed in the sodium fluoride beds. Thus, by. heating the beds, UF¢ is brought over in a very pure state, completely decontaminated and with losses less than 0.1%. The UFg is reduced to UF,; in a hydrogen-fluorine flame, and is col- — lected as & powder in a cyclone separator backed up by gas filters. Losses iafi ¢ “J - ) 1) 19 routinely are smaller than random errors in the assays, and the process has been used successfully for many years in the manufacture of enriched 235y from natural uranium in the production plants at Oak Ridge. The Fluoride Volatility Process alone is sufficient for the economi- cal operation of a molten salt converter reactor. Spent fuel containing UF,, ThF,, 233Pa, as well as fiesion products is removed from the reactor periodically and fluorinated for recovery of uranium isotopes. The = stripped salt is discarded (stored in INOR cylinders indefinitely) to purge the system of fission products. Although the discarded salt con- tains valuable components (7Li, Be, 232Tn, 23Pa), the cost of discarding these is'offset by the improvemént in conversion. ratido, il The steps described above appear to be especially attractive for integration with the reactor plant. That is, they are all high-tempera- ture, non-agueous processes,,and could convenliently be carried out in the reactor cell, utilizing the same shielding and sharing in the use of re- mote maintenance equipment. The waste product (fuel salt stripped of itotopes of uranium) is in. a form conveniently stored for decay of radio- activity. After a period measured in years, the waste could conveniently be removed to another location for recovery of thorium, lithium, beryl- lium, and other valuable components in a relatively low-level-radiation facility. The HF Solution Process (16) under study at ORNL prov1des one means of separating rare earths (whlch constltute the bulk of important non- volatile fission products, 1nclud1ng 1sotopes of samarium) from the base salt, after uranium has been remcved The separation is effected by dis- solving solidified salt in llquld HF containing up to 10% water. The rare earths, thorlum, and . related materlals pre01p1tate and mey be sepa- rated by filtration or decantatlon, permitting reuse of the. salt. The HF Solution Process is presently 1n the laboratory stage of development .2;2.15_ Thorex. Process -While the Flueride VolatilityaproceseaappeaIS;attractive if inte- grated with the reactor plant, it is not obvious that it is superior in a central facility to alternative modes of processing, such as Thorex. 20 This uncertainty is due in part to paucity of reliable infOrmation on costs of on-site and central Fluoride Volatility process plants,.and in part to the limitations of the method in respect to reccvery_of_lithium “and thorium. On the other hand, the costs of Thorex plants are rather better known, and, with.suitable modifications, Thorex appears to permit economic recovery of all valuable components of the fuel salt only mod- erately contaminéted with certain fission products (e.g., cesium). The . costs associated with a modified Thorex process as described in Section 5.3 were used in an alternate evaluation of the MSCR. 2.2.16 Fractional Crystallization Process Studies by Ward et al. (108,:106, 80, p. 80).provide a basis for evaluating the feasibility of removing rare earth fluorides from the fuel salt by partial freezing. A brief description is given in Section 6.7.3. The process is not suitable for a breeder reactor inasmuch as the fission product concentration cannot be lowered much below 0.2 mole %; however, much higher concentrations can be tolerated in a converter. In the ref- erence design studied here, the concentration is approximately 0.5 mole %. 2.2.17 Other Processes Solvents which will selectively dissolve either ThF, or rare earth fluorides are being sought at ORNL. Solutions of SbFs in HF show some promise. _ The capture of a neutron by an atom of 233Pa results in a double loss — that of the neutron and of the fissile atom of #23U that would have been formed by decay of the Pa. A process is needed that can quickly and economically remove 233Pa from the circulating salt stream so that it " may be held outside the reactor until it decays to ?22U. There is a pos- sibility that exposing the fertile stream to beds of ThO, pellets might accomplish this. There is some evidence that thorium from the beds will exchange with Pa in the solution, and the latter will be immobilized until it decays, after which it might, as 233U, exchange with thorium in the salt, and so become available for recovery by fluorination. Other oxides, e.g., BeO, are also under study. N 1) #) 21 2.2.18 Molten Salt Reactor Studies The status of the Molten.Salt Reactor Program was reviewed in 1958 for the second Geneva Conference by MacPherson et al. (56). At that time a homogeneous molten salt reactor having only a limited capability for fuel regeneration was under consideration. Further.studies of this system were reported by Alexander et al. (l) and a 30-Mwt experimental reactor was described (2). Also, in 1958, good indications were obtained that the system INOR- graphite-salt is chemically stable in radiation fields and attention was accordingly shifted to graphite-moderated systems. MacPherson et al. (60) described a one-region sihgle-fluid reactor utilizing slightly en- riched uranium and a highly enriched feed. Many features of his concept were incorporated in the present study. The potential of graphite-moderated molten-salt reactors for breeding in the Th-233U cycle was investigated and the associated development problems were identified by MacPherson in a series of papers (61-63). Several conceptual designs for one- and two-region breeders were proposed. One of these (the MSBR) was evaluated in comparison with four other ther- mal breeders by the Thorium Breeder Reactor Evaluation Group at ORNL (3); this system employed a fuel salt (contained in graphite bayonet tubes and circulated through external heat exchangers) together with a fertile salt stream (containing all the thorium) surrounding the moderated core region. The major problems associated With this concept were the development of a reliable graphlte-metal Joint for connecting the bayonet tubes to an INOR header and the uncertain: behav1or of the core structure for long periods under 1rrad1atlon at hlgh‘power den81t1es. | It was estimated that the MSER could achieve fuel yields up to about "7%/year (doubling time about 14 years) at fuel cycle costs not greater ~ than 1.5 mills/kwhr; and that fuel costs as low as 0.7 mills/kwhr could be achleved by sacr1f1c1ng the fuel yleld in favor of lower processing costs (3) o | 22 2.3 Molten Salt Reactor Experiment The favorable results obtained in the various evaluation and develop- ment programs led_to the initiation in May 1960 of preliminary design of “the Molten Salt Reactor Experiment (12,5). Construction and installation of the entire system are scheduled for completion in mid-~1964 and criti- cality late in 1964, or early 1965. The MSRE is expected to demonstrate the long-term reliability of components and the compatibility of materiais under actual operating con- ditions, including the dimensional stability of the graphite and its re- sistance to permeation by fuel salt in the presence of radiations and the maintainability of the system after operation at power. The reactor will produce up to 10 megawatts of heat in a fuel con- sisting of a solution of highly enriched 235U"F, dissolved in a mixture of the fluorides of lithium (99.990% Li), beryllium, and zirconium . having a liquidus temperature of 842°F. The salt enters a volute around the upper part of the cylindrical vessel at 1175°F and flows at the rate of 1200 gpm down through an annular plenum between the wall of the vessel and up the graphite core-matrix. This is constructed by pinning 2-in. square bars loosely to INOR beams lying across the bottom of the vessel. The salt flows up among the bars at a velocity of 0.7 ft/sec.-(Reynolds number 1000) and exits at 1225°F. ' The fuel pump, & sump-type having a bowl 36 in. in diameter and 12 in. high, is driven by a 75 hp motor and develops:a head of 48.5 ft at 1200 gpm. All parts are constructed of INCR. The heat exchanger, also constructed of INOR, has 165 tubes 14 ft long by 1/2 in. OD with walls 0.042 in. thick, and provides 259 ft? of heat transfer surface (heat flux 130,000 Btu/hr-ft°® at a IMID of 133°F). The reactor heat is transferred to a secondary salt coolant from whence it is discharged to the atmosphere in an air-cooled radiator. Initially, the MSRE will contain no thorium, since the power level is too low for significant emounts of 433U to be produced in a reasonable time. Thorium may be added later to permit verification of nuclear calcu- o~ & seen compatibility or stability problems. , » lations of critical mass, etc., and to discover if there are any unfore- 0 7} ‘densities are only 14 kw/liter of core and 35 kw/liter of salt (average). 23 3. BASES AND ASSUMPTIONS 3.1 Design Bases 3.1.1 Reactor Concept The concept.selected for study was madeled closely after that pro- posed by MacPherson et al.:(60), and is essentially a scalezup of the .. Molten Salt Reactor Experiment (12,5) plué necessary auxiliary equipment for generation of electricity, etc. Briefly, the core consists of a ver- tical bundle of unclad graphite logs conbtained in an INOR vessel. Fuel salt containing thorium and uranium flows up through the bundle into a plenum, thence through several pumps in parallel to the shell side of multiple shell-and-tube heat exchangers, and then back to the reactor. 3.1.2 Design Calculations These were performed only in sufficient detail to permit the estima- tion of the capital cost. Problems of control, shielding, hazards analy- sis, etc., were ignored. . Attention was centered on the nuclear perform- ance and processing costs. The energy conversion system was designed to provide & basis for estimating the volume of the fuel salt circulating in the primary heat system, the net thermal efficiency, and the capital investment. 3.1.3 Station Power = An electrical capabiiitjfof*ldOO,Mw was selected to permit direct comparison with systems previOQSly evaluated at the same plant capacity. Preliminary calculations indicétéd.that-the core should be ~20 ft in diam- eter for satisfactory nuclear performance. At a power of 1000 Mwe, power A lowerfplant dutput'WOuld result in inefficient utilization of the fuel inventory. 24 3.1.4 Plant Utilization Factor The standsrd factor of 0.8 was used as recommended in the "Guide" (52). 3.1.5 Thermel Efficiency Several different energy conversion schemes were considered in suf- ficient detail (see Section 4.3) to show that even the least efficient system (Loeffler boiler) would have, when fully optimized, & thermal ef- ficiency not less than 40%. This efficiency was therefore adopted for use in the fuel cost optimization calculations. 3.1.6 Fueling Cycle For the purposes of optimization calculations, it was assumed that make-up fuel was added and spent fuel was removed quasi-continuously, and that, with three exceptions, the concentrations of the various nu- clides in the circulating salt system were in equilibrium with respect to feed rates, nuclear reactions, and processing rates. The exceptions were 234U and 238U (which are initially present in amounts substantially lower than the equilibrium value, and whose concentrations increase with time) and 226U (the concentration of which starts at zero and reaches only about-3/4 of its equilibrium values in 30 years). For these three isotopes, concentrations that approximated the average over a life of 30 years starting with the reactor charged with 23°U (95% enrichment) were used. Other important isotopes appear to approach their equilibrium concentrations in times short compared to the reactor life. The use of equilibrium concentrations for these, especially for slowly equilibrating fission products, is discussed in Appendix H. 3.1.7 Processing The processing rate wés optimized with respect to the fuel cycle cost. In the selected process, spent fuel is accumulated, shipped to a central Fluoride Volatility Plant, cooled for a minimum of 90 days, and treated for recovery of uranium. Undecayed 233Pa, along with 232Th, 714, and °Be are lost in the waste. i ” zwu‘ ) 25 3.1.8 PFeed and Recyclé.b In the optimization calculations, it was assumed that isotopes of uranium recovered from irradiated fuel are recycled, and that deficiencies in the breeding ratio are compensated by additions of 95%-enriched 2357, The effects of a few feed and recycle échemes'on the optimum reactor were studied (Section 6.9), such as the use of feeds containing a mixture of uranium isotopes (e.g., spent fuel from the Consolidated Edison Reactor at Indian Point, New York). The sale of irradiated fuel to the AEC as an alternate to recycle was also investigated. 3.1.9 Isotopic Composition of Lithium It was assumed that lithium (as the hydroxide) would be available in grades containing up to 99.995% 71i at a price no greater than that quoted in reference 67 ($120/kg of lithium). The choice of this composition (rather than one having a lower cost) resulted from a compromise between cost of neutron losses to ®Li and the cost of discarding the salt enriched in 7Li with a processing rate of about 2 ft?/day. 3.1.10 Energy Conversion System Although it would be difficult to establish a complete set of require- ments for coupling of the reactor system with the energy conversion system prior to the preparation of a detailed design; nevertheless, it is neces- sary to fix some of these in order that the fuel cyéle cost may be esti- mated. The most_im@drtantrreQuirement'appears to be a necessity to iso- late the fuel salt from the thermodynamic fluid, at least when that fluid - is waterfi The hazards associated with the possibility that high_préssure P . steam might leak into the fuel system cannot be tolerated, since such .. leakage would result in the rapid formation of U0z (Sl,rp} 63). This is only slightly solublé in the base salt, althdugh its solubility can be increased. somewhat by additions of ZrF, and of ThF4 (84, p. 96). 1Isola- tion of the steam and fuel systems is achieved by 1nterp051ng & compatlble - third fluid, either as a stagnant layer or as a separate stream circulated between primary and secondary heat exchangers. 26 | The intermediate coolant'(third fluid) must be chemically éompafiible with fuel salt, and in addition, it is desirable that it be inert with respect to steam. Also, if should either not be a nuclear poison, or else it should be readily removable from fuel salt. .For the reference design, a salt 66 mole ¢ LiF (99.995% 7Li) and 34 mole % BeF, was selected (Table 3.4) as the intermediate fluid. 3.1.11 Primary Heat Exchanger Requirements It is imporfant that the external portion of the fuel salt circulat- ing system shall have as small a volume as possible in order to reduce the inventory of valuable materials. 'However, the reliability and main- tainability of the sjstem cannot be compromiséd in favor of small volume. A requirement for maintainability, which includes replaceability, implies that the primary heat exchanger shall be drainablé of fuel salt. This re- quirement is most easily and certainly metvby putting the fuel salt in | the shell-sides of the heat exchangers and grouping these about the reac- tor in a vertical position so that the heads may be removed and the tube bundles lifted out easil&. 3.1.12 Minimum Salt Temperatures To provide a margin of safety in regard to possible freezing of both fuel salt and intermediate coolant salt, it was decided that the operating temperature of any salt stream should not be at a temperature less than the liquidus temperature of the fuel salt. 3.2 Cost Bases 3.2.1 Value of Fissile Isotopes Unirradiated, highly enriched ?33U was valued at $12.01/gram of con- tained ?3°U (52). Mixtures of isotopes were valued according to the formule V = £(E) $12/gram of contained fissile isotope (223U, 23°U), where f(E) is an enrichment factor found by dividing the value of enriched 232°U having the same composition as the mixture in question by $12.0l/gram\ The enrichment, E, of the mixture is found by dividing the sum of the ifij/ it » 7) C 27 ‘atomic concentrations of 235U and 233y (and 233Pa, if any) by the sum of atomic concentratlons of all 1sotopes of uranium in the mixture (thus lumping 234U and 236U with 228U as diluents). 3.2.2 Value of Thorium Inquiries directed to several vendors elicited only one reply (Appen- dix J); however, the quotation given ($6/1b of ThF,) agreed well with a 1959 estimate by Orrosion (89) and led to the adoption of a price of $19.00/kg of thorium as ThF,; ($6.50/1b ThF, ). 3.2.3 Value of LiF(99.995% 7Li) This was taken to be $120/kg of contained lithium (Appendix J) or $32.30/kg of LiF. 3.2.4 Value of BeFo Inquiries cited in Appendix J led to adoption of a price of $15.40/kg of BeFs. 3.2.5 Value of Base Salt This varied with the composition, but the base salt in the optimum reactor contained 68 moles of LiF per 23 moles of Bng giving a value of $25.97/kg. 3. 2 6 Cost of Compounding,and Purlfylng Fuel Salt The operatlon of blendlng recycle uran1um.w1th make-up uranlum and fresh lithium, berylllum, and thorlum fluoride end purifying is to be per- formed on-site. The cost was therefore excluded from the operating and capital charges of the proce331ng plant and included in the capital and :operating charges of the reactor plant 3;2,7f INOR-8 Cost" 'The following cost information. supplied by A. Taboada of ORNL is based on quantity production. Manufacturing experience to date with fab- ‘rication of the listed forms has not indicated the existence of any 28 serious problems and therefofekpricing saféty facfiors in_the costs shown , may be pessimistic. | : o Plate | $3 per 1b Round Rod | $4.25 per 1b ~ Welding Rod - $8 per 1b Pipe (Seamless) ~ $10 per 1b Pipe (Welded) $5 per 1b Tubing (Seamless) $12 per 1b Tubing (Welded) $6 per 1b Simple Forgings $4.50 per lb‘(e.g., tube sheetS) Fabricated Plate $10 per 1b (e.g., pressure vessel shells) Dished Heads $5.50 per 1b ' Forged Pipe Fittings $50 per 1b Castings $2 per 1b 3.2.8 Moderator Graphite Cost The cost of graphite such as would be used in the MSCR core has been established at $6.00 per 1b. This is from informal discussion with ven- dors. 3.2.9 Annual Fixed Charges For fissile isotopes, the use charge was taken at 4.75%/yr in ac- cordance with the "Guide" (52). Other components of the fuel mixture were carried as depreciating assets (since only the isotopes of uranium and thorium are recoverable). For such the "Guide" recommends (Tsble 3.1) an annual rate of 14.46% for an investor-owned public utility (IOPU). This rate, however, includes 0.35% for interim replacement when the rate of re- placement is not known. In the preéent instance, the replacement rates for base salt were calculated and the corresponding costs listed sepa= rately; therefore, the annual charge for the above items was set at ... . . 14.11%/yr. This included also 1.11%/yr for amortization by means of a 30fyr sinking fund with cost of money at 6.75%/yr; hehce, & charge for replacement of salt at the end of 30 years was not made either separately or as part of the final processing to recover the uranium inventory. o it ‘._' o e 3 » 29 3.2.10 Central Fluoride Volatility Plant Processing Charges The schedule given beldw was extracted from the estimates presented in Table 5.9 and apply £0 a plant capable of processing 30 £t of salt per day (about 1000/kg day of thorium for the reference design salt) for recovery of isotopes of urénium. The -barren salt is discarded. Capital investment ($25.5 million) was estimated by scaling from a study by Carter, Milford, and Stockdale (21) of two smaller on-site plants (1.2 and 12 ft3/day), and adding costs of other facilities required in a cen- tral plant (receiving, outside utilities, land improvements, etc.). The plant is large enough to service about fifteen 1000 Mwe molten salt con- verter reactor plants. A turn-around-time of two days was allowed. Ship- ping charges ($10.30/kg thorium) were estimated separately (Table 5.8). Table 3.1. MSCR Reference Design One Ton/Day Central Fluoride Volatility Plant Cost Schedule Production Rate , - . kg/day of Thorium Processing Cost from Reactor $/kg Thorium 320 23.0 160 24.0 80 25.3 40 26.1 53.3 26.6 40 27.6 26.7 30.0 *Excluding shipping. _343_Special AsSfimptions 3. 3 1 Permeatlon of Gr_phlte by_Salt Tests with MBRE fuel salt at 1300°F and 150 p31 in MBRE graphlte showed penetratlons of the order of 0. 02% in 100 hours (86, p. 93) Most " of the absorbed salt was contained in pockets lying at the surface of the graphite, and presumably in communication with bulk liquid. From a metal- - lographic examination of thin sections, it was concluded that penetrations 30 O considerably less_thah 0.12% would be encounteréd in the MSRE at the “maximum.pressure of 65 psia. For the purposes 6f eValfiating the MSCR, it was assumed the penetratioh would beAO.l%, andkthafi‘oniy_pores iying at the surféce would contain salt. Thus, in a core 90 volume % graphite the volume of Salf absorbed in the graphite would be slightly-less than 1% of the volume of salt in the core. This absorbed salt was assumed to have the same composition as the circuléting‘stream. 3.3.2 Permeation of Graphite by *3° Xenon: , The solubility of xenon and other noble fission product gases in fuel salt is very low (107); also, their adsorption on graphite at 1200°F ap- . pears to be negligible (17). However, there remsins the possibility that gaseous xenbn may diffuse into the pores in graphite at & rate large com- e pared to that at which it can be removed from the salt by sparging or | spraying. The mathematical treatment of the case at hand has been pre- sented by Watson, et. al. (107), who also established probable ranges for the diffusion coefficient. For the purposes of a reference calcula- tion having a reasonable degree of plausibility, a value of 10-6 (cm?/ sec) was selected for the diffusion coefficient and a value of 0.0l for the porosity of graphite to noble gases. Further, it was assumed to be feasible to by-pass 10% of the fuel salt (16 f£t3/sec) through the pump bowls or through a sparge chamber, and that this by-pass steam would give up substantially all of its xenon to the sweep gas. '3.3.3 Corrosion Products Tests in a forced convection INCR loop using a salt (62-LiF, 36.5- o BeF,, 0.5-UF;, 1.0-ThF,) very similar (except for thorium content) to that proposed for the MSCR,'show:that after a period of initial attack (occurring generally in the equipment in which the batch of salt is pre- pared) the concentration of structural-element cations reaches equilibrium values (84, p. 79). The temperature of the salt was 1300°F in the hot leg, 1100°F in the cold leg, and was circulated for a total of almost 15,000 hours. - The concentration of nickel, after rising to a‘maximnm-of‘ o~ 80 ppm in about a thousand hours, reached an equilibrium value of about '_\,J o s »n values. 31 50 ppm at 2000 hours. _Chromium-concentration fluctuated between 400 and 600 Ppm, .averaging about 500 ppm, while iron averaged about 250 ppm. Molybdenum was said to be negligible and was not repbrted. Apparently thegconéentration of chromium is in equilibrium with respect to the .rate with which chromium is oxidized by UF4 to CrFo at the hot metal surfaces and the rate with which it is reduced to O¢r at the cold surfaces (75, p- 39). In the MSCR, large areas of INOR are exposed to the salt at all temperatures between'llOO°F‘and 1300°F. Although the rate of diffusion of chromium'in INCR has béen determined at various teriperatures, it is not possible to calculate the chromium concentration in the salt until the temperature profile is known. | In the calculations performed here, a neutron-poison allowance was ~made for corrosion producf?, amounting to 0.008 neutrons per atom of fuel destroyed. This loss is comparable to the loss that would result if the concentrations of Ni, Cr,_and Fe were 50, 500, and 250 ppm, as in the loop-corrosion test cited above (Section 6.3). 3.3.4 Approach.to Equilibrium The nuclear performance was calculated by means of MERC-1, an equi-’. - librium reactor code. Thus the performance of the reactor during the approach to equilibrium, when concentrations of isotopes of uranium and of fission products are changing, was not considered, except in regard to 234y, 2367, ‘and ?38U. These were averaged over a fuel lifetime of 30 years; 223U, 235U -and flSSlon products were taken at their equilibrium In cases where adequate supplles of ?33U are unavailable the reactor ‘would be fueled initially with enriched 235U. This is inferior to 233y ‘in respect to eta.and aISO-formsra non-fertile daughter, 226U. _These ‘disadvanteges are offset by.initially-low concentrations of fission ) | products and 236U. While a calculatlon of the time-dependent behav1or is 'de81rable in such cases, 1t does not appear that the error 1ntroduced by | assuming equilibrium pondltlpns ;s important. The matter is. explored further in Appendix H. 32 4. DESCRIPTION OF MSCR CONCEPT 4.1 General Description The MSCR is a single-region, uhreflected, graphite-moderated,fluid- fuel reactor utilizing_a mixture of molten fluorides-of lithium, beryl- lium, thorium, and uranium as the fuel and primary coolant. A sketch of - the reference design reactor is shown in Fig. 4.1. As seen in this fig- ure, the reactor consists of a 20-ft-diam by 20-ft-high cylindrical core made up of 8-in.-diam graphite cylinders. The fuel salt enters through a bottom grid, flows upward through the spaces between the cylinders- and is discharged into one of eight primary heat removal circuits located around the reactor. The arrangement of these circuits is shown in Fig. 4.2. The heat generated in fuel salt is transferred to an inteifiediate coolant salt consisting of a mixture of barren lithium and beryllium fluoride containing no uranium or thorium. The coolant salt is used to superheat saturated steam produced in a Loeffler boiler and also to re- heat steam from the turbogenerators. The reactor vessel, internals and all primary and secondary system components in‘contact,with”fuel salt and coolant salt are constructed of INOR-8. The specifications are tabulated in section 4.10. Part of the superheated steam is sent to & high-pressure turbine and the rest is injected into the Loeffler boilers to generate saturated steam. This saturated steam is recirculated to the superheater by steam-driven axial compressors using steam drawn from the‘high-pressure turbine discharge. A flowsheet of this heat removal-power generation sys- tem is shown in Fig. 4.3. ‘ - Design data and operating characteristics for the reference deéign are given in Table 4.1. 4.2 Site Plan The site plan of the MSCR plant is shown in Fig. 4.4 based on con- ditions specified in the AEC Cost Evaluation Guide (52). The 1200-acre grass-covered site has level terrain and is located on the bank of'a river. Grade level of the site is 40 ft above the river low water level P : 33 ORNL-DWG. 63-2801R1 MOTOR SALT LEVEL CALANDRIA . L H PUMP jT=—f= —— FReeZE |f - ”HHu ] ] PRIMARY HEAT VALVE SRR H = EXCHANGER COOLANT : —D — “EH” = . - SALT - g |3 Z ;{;1!’11 ——— 2 ’~== /4 23( ] = = £ £ : 1A } | 3 | ] VESSEL = 36’ | ANNULUS I - Il I GRAPHITE N T CYLINDER - CORE . FREEZE - = e VALVE 4 FUEL SALT _ f: RE e e —_— 7 &‘, SUPPORT ‘ "‘ . o - . ‘ N-PRESENCE OF LIQUID IN | ' TYPICAL CHANNEL INDICATED ‘ _ BY. HORIZONTAL LINES - 20‘ . Fig. 4.1. MSCR Vessel and Heat Exchanger. i = REACTOR VESSEL ORNL~DWG. 65-7906 8. HTR; ) } / \\ PRIMARY HEAT EXCHANGER ( { ‘ FUEL PUMP | I i | — COOLANT PUMP ; : | l @ @ SUPERHEATER 1A (] | Ene. | I i r - . i _ = \ \] fl’/ \ REHEATER | N 1 e ~ \\ R d - // — 9 S { ' A %' -~ / . NS \ , p S \\ M N — - —— /L , \ { _ e e T 2 — 3 \\ COOLANT SALT LINE @ ‘ = -— N\ 12°8eh, 20 SHIELD Fig. 4.2. MSCR Heat Transfer Flow Diagram. e 35 # FREEZE VALVE _ CABLE WINCH : SIS ORNL-DWG, 65-79C LIF BeF, RA . LI T | & 1] S it - @ DRY CHEMICAL £ SHIPPING PORTABLE | .% N FEEDER } g g i coouNT N ) : : Wyin rsr? IS T - ”FF:Q"DL','"'W ) % ] oay _ 2-3Y x 1% in. wALL : sssxifw. | S oA D ; T A1 cvemear {‘ soaxctr - 845p, 1000°F, 1520 } STATION SUMMARY ¢ L SALT ADDITION ~———— 'g FEEDER -g | r..;........;%:"’.‘l‘l‘ic.“_“_" _______ e Pt == 755 i REACTOR 2500 Mot —~—— " S e Ty ' ' emsemmn____ feasdhewn || | aeamenaon 103 Mo - ) | L 2 Eony b—tlyin. ' {2455 p, 1000°F, 1460.5 2400p || 8.04X10% 1 ’ | OVER-ALL PLANT EFF. 41.5% 5 tn |y METER -hp | : | | i ™ PUMPS COOLANT | AUX. POWER 45 My /Sovsod su'-ao N son’] | fuEL P\ | l | | 4 > - _ PN PUMP ' : t-in. SCH.40 T . R 240 SCH. 8-2000 hp ‘ | |re - TLINE ! $2-in SCH40 I | Xxpe/ = * ' ! ! | . L % METERNG R | l e - ! 2n’ 00 L o | #-n | | % = | » gem b 1seH.20) ! o | 200 [ 1o X REACTIVITY : | TEMP wscr | soraoU)- controe | | Rasx® | § .. DRaN | | 8- x 2Y%in. i ! > - 2-500° | guoer | wau ' = ' ‘ . ‘ 350p : MELT CHEMICAL . rzm\m SALY : / L - | 1-12-in. SCH. BO TANK TREATMENT SALT '\ STORAGE . o s Lt scnzo {10R136.000 triv | === 15013 | 1] vamx 100 #® STORAGE \ TANK, 5-650 1> \_ ot > T ' | TANK ) p, HOOF j 250p, HOO*F jos27x0® T \ ' % : 35k R . {-mSCHA0 4-2-1n. 5CH.40 } — _ru-n-nsuueo_} | { / llu-nscu) ' . - s W e i . REAGE"'O‘? GAS ! 3in. SCH. 40 wudl : . ID- | g-44- SCH.60 m-:-‘-;-x-» b+--+ —1——-! -— ' * N ' 300p ; TN 2400p, HOIh ' ~="T"== |orP DISPOSAL AREA | | E: | - rinsonao | 50,000 1b ' t e ' | IE E ' ' START-UP | ] - | l_l = ! BOILER souer| ) | 180p FLASK TRANSFER = oI ARED) 11 7 wny TANK DECAY STORAGE TURBINE ’ Py d 03 TANKS, 5-35 13 " cHEMICAL COOLANT Sall " 34 0 NKS, o Tosm \ : ELT TANK TMENT DRAIN AND STORaGe STEAM . y1ls L jossaxcte . . . : m 2‘m "3 uRc‘ ' ¢ . o - _J -— FROM REAGENT | COND. TANK, 50 , ! i} . SHIELDED SHIPPING ] ,r 0 l‘. W15 1619 (F z CONTAINER gl REAGENT GAS 1 L * '3 FUEL SALY DRAIN DISPOSAL AREA 18-In. SCH.160 i} 5-18-in. SCH. TANKS, 54-80 4> 8.04% %% 1 if] 00t sz 264 m9ver ' LIUID WASTE = R FROM ——tme S4S°F, s42.4h, 1|} | P P p : ‘ ' REAGENT GAS 25009 } 1 SUPPLY . i H1 : | " i He - I 18th | | | IR, . A ol A A L.t - s o oo el SEES Ane Sute dmms GREE JENL SUNAD G SEID SEN S Sm—. J Fig. 4.3. MSCR Composite Flow Diagram. RIVER FLOow 36 2¢g "] 375 ———" st A —————— 100000 GAL HEBATIMNG '/ O BTORALE TANL —t SBal. WLl | =¥ A ¥ -96S « 2 | — — DO e, S e e ee o 10 TR AVELY | ISP TIDIRA0- SeReanalF) 1P - } b - b b » F4———FF] b s YRm . BUMPS SCEREM WA S AR, ¥ £l pusip. + SERVICE WATER DUMP S, | CURE. WATER | WITAK.S PipEL 2-f00$00 GAL. . | remas r - Takis 9. e ol GAle. HEATRMG oy TaM.. ~ N, ) WALTE SibPeBAL 1 sonog 1 | t::::* . - _——] i sarrrrly e 1 ——— ——— m— — ¥ i ¥ — e ¥ it rarrr———— DREP WaLL Perap —-—¢- Funce) N Py )k waSTE €Al T i . Ty FELEG HOILBRS, 1 T [ S [ L] N N ety — oG \‘ N M jvh_t‘a_: e zasui. srss R O A D ‘ o ¢ S WITCHYARD P-rasawes Q - n ey —y N Fig. 4.4. MSCR Plant Plot Plan. (Sargent and Lundy Drawing) » Table 4.1. 37 Design Characteristics of the 1000 Mwe Molten Salt Converter Reference Reactor General Thermal power Net thermal efficiency Net electrical power Core geometry Moderator Form Dimensions Weight Volume fraction in core Porosity accessible to salt (assumed) Porosity accessible to gas (assumed) Gas diffusion coefficient (assumed) Graphite density " Radiation heating (max.) Maximum temperature rise Reactor vessel Inside diameter Thickness Maximum temperature Weight, including internals Radiation heating in support plates Radiation heating in vessel wall Maximum temperature rise in wall Fuel stream Composition Base.salt (LiF- Bng-ThF4) _ 2500 Mw 41. 5% 1038 Mw Cylindrical, 20 ft X 20 ft Unclad graphite Cylinders 8 in. diam, 24 in. long 335 tons 0.9 0.1% 10-¢ cm?/sec 1.9 g/cm 5.2 watts/cm’® 520°F INOR-8 20 ft-2 in. 1.7 in. 1400°F 125 tons 2 watts/cm’ 0.6 watts/cm’ 40°F 68-22-9 mole % UF, (fissile) 0.3 mole % Fission products 0.5 mole % Corrosion products 750 ppm quuldus temperature of base salt 887°F Density of base salt at 1200°F 3.045 gfcc Mean heat capacity of base salt at 1200°F 0.383 Btu/1b-°F Fraction of core occupied by fuel salt 0.1, Fuel stream inlet temperature - 1100°F . Fuel stream cutlet temperature 1300°F ~ Flow rate 160 £t3/sec Velocity in channels in core (avg ) 6 f£t/sec - Velocity in piping | 35 ft/sec Velocity in heat exchanger Shell side 20 ft/sec Tube side 31 ft/sec 38 Table 4.1 (continugd)r Fuel Stream (c¢ontinued) Pressure, psia Pump discharge Heat exchanger inlet Heat exchanger outlet Reactor inlet Reactor outlet Pump suction Power density in fuel salt In core (max.) Average over entire fuel volume Volume of fuel salt In active core In top and bottom plena In fuel annulus adjacent to vessel wall In surge tank In pumps In heat exchangers In connective piping In dump tanks and reactivity control tanks TOTAL Volume of fuel in active core Primary heat transfer loop Primary pumps; number and type Pressure at pump discharge Primary heat exchangers Total heat transfer area Average heat flux Material Weight Secondary heat transfer loop Coolant salt composition (mole %) Coolant salt inlet temperature Coolant salt outer temperature Coolant salt flow rate Coolant salt pump discharge pressure Coolant salt volume 190 185 95 80 35 22.5 510 w/ce 35 wfce 630 ft2 540 £t 105 f£t3 85 £t3 130 £t 575 £t2 320 ft3 115 ft? 2500 £t> 650 £t3 8 — Salt Lubricated 200 ‘psi 8 — Shell and Tube 53,000 ft2 160,000 Btu/hr- £t2 INOR 8 36,000 1b each 66-LiF; 34-BeFp 950°F 1100°F 203 £t>/sec 350 psi 5600 ft> FL O » Teble 4.1 (econtinued) 39 Energy conversion loop’ Superheaters Materials Heat flux Heat transfer area Weight Inlet steam temperature. Steam flow rate Reheaters Materials Heat flux Heat transfer area Weight (approx.) Inlet steam temperature Steam flow rate Loeffler boilers Length Diameter (ID) Weight (approx.) Inlet steam conditions Steam flow .rate Inlet feedwater conditions Feedwater flow rate - Discharge steam conditions Steam circulators Flow rate Power Steam temperature Steam pressure Turblne Flow rate , Generator output -Steam temperature - Steam pressure '_Exhaust Proces51ng system | Processulg method | Salt processing rate - - Production rate Cooling time. (average) 16 = U-Shell and U-Tube INOR-8 and alloy steel 52,440 Btu/hr ££2 8, 850 2 ~50 000 1bs 670°F 1.3 x 108 1b/hr 8 — Shell and U-Tube INOR~-8 and alloy steel 37,250 Btu/hr-ft? 3, 543 ft? 22 000 1ibs 635°F 0.7 x 10° 1b/hr 4 100 ft 6 ft 600,000 lbs 2430 psia/1000°F 3.1 x 106 lb/hr 2520 psia/545°F 2.0 x 108 lb/hr 2400 psia/662°F (sat.) 4 (turbine driven) 20.5 x 10® 1b/hr 5,100 BHP 670°F 2,480_p$ia 1 (CC6F-RH) 8.04 x 10% 1v/hr 1083 Mw '1000/1000°F 2400/545 psia - - 1.5 in. Hg Central Fluoride Volatlllty 1.67 £t3/day 53.3 kg Thorlum/day ~90 days 40 Table 4.1 (continued) Processing system (continued) o Hold-up time (total) o 116 days Processing batch size ~6,000 kg Thorium Processing plant capacity 1,000 kg/day Turn-around time ' ' 2 days ' and 20 ft above the high water level. An adequate source of raw water for the ultimate station capacity is assumed to be provided by the river with an average meximum temperature of 75°F and an average minimum tem- perature of 40°F. 4.3 Structures' Plan views of the reactor and turbine building are shown in Figs. 4.5 and 4.6 and vertical sections in Fig. 4.7. As seen in these figures, the reactor building and turbine building are adjacent, the secondary shield wall forming a separation from grade to the main floor. The buildings are two-level structures with the grade floors of the turbine and reactor buildings at an elevation of one foot above grade, and the main floor at 36 feet above grade. The secondary shield wall extends to the main floor and forms the walls of the lower part of the reactor end auxiliary build- ing. | The turbine building and the upper level of the reactor and auxiliary buildings are steel frame structures, with insulated metal panel siding. The arrangement of the equipment within the buildings is indicated on the general arfangement drawings, Figs.. 4.5 and 4.7. ' A three-level steel frame and insulated metal-panel structure ad- joining the turbine building houses the administrative offices, control room, switchgear, batteries, plant heating boiler and makeup water de- mineralization plant. Lockers, showers, and toilets for plant personnel are also located in this building. | A 200-ft waste gas stack is provided for dispersal of plant venti- lating air and waste gases from the various reactor and reactor equipment rooms. 41 1A_I e LOE ML ER COOLANT \ 1 REMERVE AL, TRANSE, uNT AN, TRANDE L. ™, TykBina A B . e TURBING TRANMSY. :EJ Fig. 4.5. MSCR Plant General Arrangement. (Sargent and Lundy Drawing) UNAT TRANLF, IECULATING WATER DISCHARGE TUNNEL + e e Y8 A ; ; CARECULAT SQLAMT A%, sTRIPPREY ALT PUHMP S vaLYuM AR BIECHARGH TILTRE, AL CHMR. NE B CYELE CorPRaL S SoRe coNe - WALTER TAMK waten, cfim. SEPISTRE CoNDENLER raPSRATSR U CRUPRNARR PUMmSY HE CHAR, ADS DRMCAY Toain, VACULM TANK FLoo® wL. 9860 r M oo (5 0dd Ghk. RAT FREADWATRR bumpPs -1 DRABRATOR PPN G MDD DI TPV DA FRRIAUT RESERVE AUX, TRAMNSFE. URET AUR LB TURBING UNIT AUKR, TRANSF TRANSFORMER TRANSF, H.P TURBINE TRANSFORMER " c- Fig. 1.6, MSCR Plant — Plan. (Sargent and Lundy Drawing) 43 120 TON CcRAN® 150 TON CRANE Lu oo “5“;! *II-I: ; ;’: T @t 3 + o osa%a’ *l.. 10800 cooLANT ‘ . 10%0-a" Pruraa, EL.joyo-¢ saete” , ' .. 908ie’ - $88L0" TaNES Leal. 984" SECTION AA CTION "B-&’' tankAToRd S Nes e 50 TOM CRANR L. L A G ContRo. Roam ¢ GEW), OFrican CamLE Room { "TCHARAR ERROwWATER e el - - - e e - o Pamps =, 0ae! . o s o — GEAD LL. Dagl o Cilke, wWATIR Dinew Tunnte SECTION "c-¢" Fig. 4.7. MSCR Plant — Elevation. (Sargent and Lundy Drawing) bt 4.4 Primary System Components The primary system components consist of the reactor vessel, modera- tor, fuel salt pumps and fuel salt-to-intermediaste coolant salt heat ex- changer. The arrangement of these components is shown in Figs. 4.1 and 4.2. Design and performance characteristics are summarized in the follow- ing paragraphs. 4.4.1 Reactor Vessel The reactor vessel is 20 ft in diameter and 37 ft high (including expansion dome) as shown in Fig. 4.1. The vessel wall is fabricated of 2-in.-thick INOR-8. No thermal shield is required since the gamma heat- ing of the vessel wall does not exceed 0.6 w/cc. The thickness of the external vessel insulation and salt flow through the core-vessel annulus can be adjusted as necessary to avoid excessive thermal stresses in the vessel wall and at the same time minimize heat loss to the external con- tainment space. A 6-ft diam expansion dome at the top of the reactor vessel not only provides surge volume but serves as the fuel salt volatiles purge loca- tion, the UF, pellet injection point and salt sampler location. Circula- tion of fuel salt in the dome is accomplished by recirculation of salt from each pump discharge through an orificed 2-in. line to a point below the normal ligquid level in the dome. This level is maintained by adjust- ing the pressure of the helium cover gas. To minimize holdup of salt, 50% of the volume of the dome is occupied by 2-in. diam sealed INOR-8 tubes. The reactor vessel is provided with eight inlet nozzles which dis- charge fuel salt radially into the bottom plenum, eight outlet nozzles leading to the pumps, together with bottom and top INOR-8 grids for sup- port and restraint of the graphite core. These grids in turn are sup- ported by columns or stanchions attached to the reactor vessel. The reactor vessel specifications are given in Section 4.10. 45 4.4.2 Moderator Structure The graphite matrix is composed of cylindrical logs & .in. in diameter and 24 in. long, as shown in Fig. 4.8. These are stacked in a vertical position and alligned by means of axial pins and sockets. Fuel-salt flows up through the cusp-shaped passages between logs. The "pile" is centered in the vessel by metal pins protruding from the support plates at the bottom and top of the vessel. These are located near the axis of the vessel and mate with a corresponding moderator log. Initially the pile rests on the lower support grid; as the reactor is filled with salt, the pile floats up against the upper grid. The pins, while allowing this vertical motion, keep the central logs centered in the vessel. The remaining logs are bound to these by means of metal hoops passing around the peripheral logs. These hoops are fabricated from mo- lybdenum, which has about the same coefficient of expansium as graphite. This arrangement ellows the support grids to expand independently of the moderator as the reactor is brought to operating temperature. Also, the increase in height of the vessel on expansion is accommodated. The radial profile Of temperature in the fuel-salt is flatfened by proper distribution of the flow, which is accomplished by orificing the flow channels. The bottom row of moderator logs is machined from hexa- gonal pieces. A 4-in. section of the end in contact with the support grid is not machined. If close-packed in a triengular lattice, these ends would block complefiely the flow path of the fuel salt. Therefore, the corners of the hexagons are cut away to provide orifices of appropriate diameter for each channel. bob.3 Fuel-Salt Circulatlng Punps Clrculatlon of the fuel salt flow1ng at 9075 gpm is maintained by a , centrlfugal pump as shcwn in Flg. 4.9 in each of the eight independent ~ heat exchange c1rcu1ts. Because the system hes no valves or other means of equipment 1solat10n, the pumps are 1nstalled at the hlghest elevation (and the highest temperature region) of the circuit between the reactor and the prlmary heat exchanger. By this means it is pOSSlble to avoid the hazard of seal flooding and reverse rotation at standstill and to :. 46 ORNL-LR-DWG 63-2803 Reactor Vessel ‘Insulation 0 down ‘ T fit..\ Wall dgr A 7l i g 7 \\.M.\.‘...\\\\ 1/ 1 f .\s..\....m ? . r L \..u... \W. \_..m“ \\..\m \\ \\ m\ ,.\K\ “ Core Hold- Grid / i 7)) Z Pump Suction Nozzle L Dome \\\.. _\ Inside Diam. g"diam. Graphite —* Cylinders ZSult Annulus Graphite Brick MSCR Core Configuration. Fig. 4.8. OIL-LUBRICATED DOUBLE- ROLL BALL BEARING HELIUM PURGE tN ~—~ SALT-LUBRICATED HYDRO- TR DYNAMIC JOURLNAL BEARING N b AN 7 PrEsid vowTe~8 — - S ' ~ IMPELLER Fig. 4.9. Molten Salt ORNL-LR=-DWG 417654 / MOTOR OIL SLINGER FACE-TYPE SEAL —» SEAL LEAKAGE OUT COOLING BARRIER HELIUM SALT LEVEL Journal Bearing Pump. 48 | o . - minimize some design problems related to thermal expansion, shaft seal % and guide bearings, and maintenance accessibility. Locating the pumps SO close to the reactor, however, introduces the prOblem‘that.the'organic materials used in the motor winding insulation and'bearing lubricant must be shielded from radiation. Also the pump motor and oil-lubricated bear- ings must be located at some distance from the salt region. In the design shown, the motor and oil-bearings are positioned above the pumps for easy maintenance and a lower salt lubricated bearing providéd to take the radial thrust. ' & by a Z500 Mt 1 / loracm/b/br > Melt ? o f#* N 5-6504¢? IIO.P-:O"M/I. / ,f;"?.';" - . - 0 - o . Tank Shamcal, Ferbie 5ol \ Flush Salt =755, 1100 / /50/33 Fande Sterage Tank ]’fordfl' " 10"schlo ¥ 7e . 100f ¢ ants PP Dy /s“ukw Reageot Gas fchte % ’ Dispesel Hrea v- I fl.k 1of? E E Coolapt Jalt Jhipping ? Melt Chemical Orain & Storage rom Flask rfi?.,ff r Tan# s ||Teatmet 2-7504(7 Reagent Cas 5-35f(% Decay S 150K °"§‘ ¥ Supply Séorage Tanks Con fo o /?capanl' Cas, Prsposal Area : From 8¢ - 50F¢ ’ Fuel "Salt of‘mgcnt Gag Supply Ovamn Tanks tobinun ohiqud W ;‘:ahm ¢ Fig. 4.13. MSCR Fuel Handling Storage Diagram. 6l . | flush-salt storage aree,'is designed for melting approximately 7,ft3/hr. Molten salt is transferred by pressure siphoning. Solid salt is added by means of a tube extending from the floor above the storage area to the melt tank. The top of the tube is flanged and sealed. A valve is located just below the flange. The tube takes a devious route and is provided with a helium purge to prevent back-flow of radiocactive material and intolerable radiation levels at the open feed point. A check valve just below the flange allows purge gas to be di= rected into the melt tank. A portable hopper is connected to the flange face to receive salt from the shipping containers. A portable shed covers the work aresa to prevent the spread of toxic dust resulting from the feed operation. 4.8.3 Coolant'Salt'Preparation Equipment and system,deSign:are similar to that of fuel-salt melt tank and solid salt feed. .The melt tank is located in the heat exchanger room near the chemical treatment tank. Solid salt is fed from the floor above. . 4.8.4 Reactor Salt Purification The reactor fuel and flush salts are purified in the molten state prior to charging or storing. This is done initially to remove oxides, and subsequently, to remove oxides and other contaminants. _Invnormal bperation,'afchemical treatment tank of 100 cubic ft ca= 'pacity located inrthe fiush'seltsstorage area receives the molten salt. After a one-day holdup for purlflcatlon, the salt is: transferred to stor- ‘age tanks. By the addltlon of L1F and - ThF, to flush salt ‘in the melt 'tank fertlle salt is formed thls is then transferred to the fertlle salt ;-purlflcatlon tank where gaseous HF and Hg are bubbled through ‘the liquid to remcve oxides. Subsequently, the treated salt is transferred to the 100 cublc It salt-storage tank 1nstalled at the same locatlon. ‘Flush-salt make—up is processed in the same manner, and- through the same equipment. - After treatment, it is transferred to a spare flush-salt tank. For the initial charges of flush sgalt and fertile salt for the reac- tor system, greater purifying capacity is required in order to avoid delay 62 in preoperational testing and‘power production. To fulfill this condi- tion, HF and Hz bubbling facilities are providéd in the fertile salt stor- “age tank -and temporary flanged-lines-arranged so that the fertile salt tank may receivé molten salt from the melt tank, perform paertial purifi- cation and allow transfer of the salt to the normal treatment tank. In this manner the two tanks, fertile Storage and chemical treatment, are placed in series to double both the purification rate and the system salt charging rate. This procédure does not increase the sampling or analysis requirements over those necessary for normal operation. 4.8.5 Coolant-Salt Purification The purpose of the coolant salt purification system is to remove impurities such as corrosion products or oxides which could cause fouling - of surfaces and plugging of lines and tubes if allowed to accumulate. . - It is unlikely that a single coolant. charge could be used for the whole lifetime of the reactor without exceeding permissible concentrations of oxides or corrosion products; however, it probably will not be neces=" sary'flaembloy a continuous treatment of coolant salt. Oxides may get into the coolant salt by accidental exposure to air or water vapor, or from oxygen present as an impurity in the cover gas used to pressurize the | coolant-salt systems. Although the rate of corrosion of INOR-8 surfaces by coolant salt is low, the area of metal surface in contact with éoolant salt is very large, so that corrosion products are certain to accumulate. When it becomes necessary to repurify the coolant salt, it is done batch- wise, one coolant circuit per batch, at infrequent intervals during peri~" - ods of reactor shutdown. If the coolant salt should'bécofie'contaminated‘with fission products or uranium as a result of a leak in the primary heat exchanger_(an event not likely to occur because the coolant salt system is kept at a pressure higher than that of the fuel salt), the contaminated éoolant_salt is .drained from the affected circuit to a drain tank, and then transferred to the chemical processing plant for disposal. th L1 4 63 4.8.6 Reactor Salt Charging System Fuel salt without uranium is injected into the primary reactor cir- cuit by manual control from a small salt addition tank to replenish the fuel salt withdrawn for chemical processing (two cubic ft per day). This addition is made remotely and with assurance that fuel salt will not flow back into the addition system. The make-up is fed into the reactor in a molten cOnditien,by gravity flow or under pressure. Sufficient electrical heating to maintain the salt in a molten condition prior to injection is provided. o A two-cubic-ft salt addition tank (metering tank) is located on the wall of the biological shield at a level somewhat above the liquid level in the reactor. The tank iS'supplied from the fertile-salt storage tank situated in the flush salt Storage.area, Makeup is accemplished by open- ing the freeze valve and the vent valve'between the tank and reactor dome. The freeze-valve in the line leading to the reactor domeé must be placed at the bottom of a loop so that when the tank is empty a heel of salt will remain in the valve. Molten coolant salt is injected into any one of the eight intermedi-: ate coolant loops from the coolant drain tanks at & rate up to 600 ft3/hr. The drain lines from several locations in each of the eight intermediate coolant lodps'also serve as charging lines for this cperation. 4.8.8 UF, Addition Facility During steady-state operation‘at equilibrium, about 6 kg/day of UF4. are added to the reactor to cowpensate for burnup and a fuel w1thdrawal of 2 ft3/day., The amount and rate of addltlon of UF,; to the reactor is governed by react1v1ty and temperature reqnlrements. Valves 1n the UF, addition system.must be kept gas-tlght when closed even though they must pass SOlldS when opened but s1nce they Wlll be in a region of relatively low radlatlon level, plastlc seats (for non-scorlng propertles) may be used. 64 The UF, charged into the reactor system is prepared byfimixing re- cycled urenium with fresh uranium. Pellets about 0.75-in. in diem, fab- ricated by casting UF4 in an inert atmosphere, are charged 1nto 8 gas- tight shielded shipping container, contalnlng an atmosPhere of dry helium at approximately 1 atmosphere absolute pressure. At the reactor, the shielded shipping container is mounted on the fuel charging machine (whlch 1e located above the reactor), and the mated assembly of the shipping con- tainer and fuel charging machine is made gas-tight by bolting a gasketed flange. Figure 4.l4 shows a schematic sketch of the UF,; addition facility. 4.8.9 Fuel Salt Drain and Storage System Facilities are prov1ded for the drainage and storage of fuel salt during periods of maintenance and emergency shutdown. The system is shown on the flow diagram in Fig. 4.2. All salt circuits have drain connec- tions; this includes the primary heat exchangers as well as the reactor vessel. The draining of the fuel salt is accompllshed quickly, even under emergency condition without steam flow and auxiliary power, since: without heat removal the salt temperatures rise 300—400°F in the first half hour. Four equally spaced 2-in. drain lines connected radially to the bottom plenum of the reactor vessel, with l-in. interconnections to the primary heat exchangers drain connections, will drain the reactor system in apQ proximately one-half hour after the freeze valves are opened. The tanks are capable of receiving the full inventory of the reactor within half an hour, maintaining the.etored volume of salt at e tempera- ture of less than 1400°F, and recharging it to the reactor. Gravity drain and gas pressure are used for transferring salt. Because forced circulation may not be available under emergency con- ditlons, natural convection cooling is provided in small-diam cylindrical ‘vessels.‘eFifty—four_vertiCally_mounted 35-in. ID cylinders 8 ft long are located in a 10=ft trench around_the inside periphery of the reactor con- tainment cell. Each cylinder is irmersed in a molten alkalil metal car- bonate bath. The carbonate mixture is contained in a double-walled tank which serves as a boiler for the removal of decay heat from the fuel salt. s Fifteen-kilowatt electric heaters are arranged within the ¢carbonate mix- NG ture annulus for fuel-salt melting following long: storage periods and. 65 CRNL-LR-DWG 75326 Shielded Shipping - Container UF¢ Pellet Container — |- .« Flange with Soft Metal Gasket Helium Purge—-[)fl— Gas Vent -—-|><|-—‘ <«—1— | cu.ft. Pellet Bin Helium Purqa—n{)fiJ . and Dispenser Gas Vent -—M—-l @ Helium P _.N_J X Metering Chamber u urge o A UF‘ P&”Gf Gas Vent «D<}— | _ ® A | Flange Disconnect Top of Reactor Shield | _ ' ] Maintenance Valve WL, UF, Addition Pipe Vent Opening , - , Reactor Dome Gas Space Salt Levgl Perforated UF, Dissolver Tube Fig. 4.14. MSCR UF, Addition Facility. 66 temperature meintenance during shorter periods. The heaters are sized to bring the salt to 1100°F in four days. 4.8.10 Coolant Salt Drain System These drain tanks have several purposes: (a) to provide storage L space for the complete volume of coolant contained in two (ofit of eight) coolant circuits during periods when it is necessary to drain the coolant from one circuit for maintenance; (b) to provide the reservoir. and the application of motive force for transferring coolant salt into the reac- tor coolant circuits during initial‘salt—charging; and (¢) to serve as a transfer point from which coolant salt may be removed from the system and transferred to shipping containers in ‘case it should become necessary to process contaminated coolant salt. The drain tanks are located at an elevation which permits the coolant salt to flow by gravity from the reactbr coolant system into the drain tanks. The transfer lines are valved with freeze-valves to permit flow: to or from any one of the 8 coolant circuits. The transfer lines are large enough to fill one coolant circuit in one hour,cusing a gas pressure of 50 psig. 4.8.11 Spent Fuel Withdrawal System Fuel salt is drawn from the reactor vessel drain line at about 30 psia into the vented metering tank. The metering tank is a 4-in.:(ID):by 25-ft-long cylindrical vessel located at an elevation such that when flow stops it contains two cubic ft of fuel salt. Vent connections above the salt in the tank lead to the dome on the reactor vessel. Decay-heat re- moval is accomplished by radiation and convection to the reactor contain- ment cell atmosphere. Because the heat generation rate diminishes rapidly, the rate of removal is controlled by means of an insulated jacket 4 in. thick surrounding the tank, and separated from it by a small air gap. A temperature-actuated bellows opens the jacket to allow excess heat to be radiated to the room. A 12-kw électric heater provides pfeheating and aids in maintaining desired salt temperature when the tank is full. After a 24-hr holdup, the salt is released by gravity drain to a transfer tank. When a 10 £t batch of salt is. accumulated, the salt is T " 67 transferred to a shippihg'flask, which is placed in a shielded cask and shipped by rail to a central processing plant. Spent fuel is accumulated for about 120 days at the processing plant, and'is then processed in about 6 days. 'Should it be necessary or desirable to hold spent fuel at the reactor site (Section 5.3.4), five 35 ft3 intermediate storage tanks are provided. Heat is removed by boiling-water in steam jackets; the steam generated is condensed in water-cooled condensers. 4.8.12 High level Radioactive Salt Sampler The high level radioactive salt sampler shown in Fig. 4.15 is used to obtain samples of fuel or flush salt from several points such as the reactor (via the fuel withdrawal tank), the transfer tank in the shipping area, the fuel drein tanks, or the flush-salt chemical treatment tank, and to deliver these samples to an adjoining hot cell for chemical analysis. Sampling is accomplished by means of access ducts located at appro- priate points. These are mounted vertically with no pitch less than 50° from the horizontal. They are large enough to allow the free passage of the sample capsule, and are equipped with a sample capsule cage at the sampling point to limit the depth in the salt melt reachéd by the capsule and prevent the end of the cable from dipping in the salt. The thief sampler principle is employed wherein an open sample capsule made in the form of a thick-walled, round-bottomed bucket hanging on a flexiblevéablé is loweréd into the salt—containing vessel. The cap- sule sinks below the Surféce.and'is filled with salt which solidifies.when the samplé capsule islwifihdrawn.td'alcooler region above'the sample'point- After solidification, the capsule is pulled up through the sampler access 'duct_into the sampler cavifiy'in which is located the cable drive mechanism. Once in the sampler cavity, the capsule is positioned over a second duct leading to a hot analytical cell. The sample capsule is lowered through the‘Second duct .and qepositéd;in'the hot cell. Usifig_a suitable manipu- lator, which is part of the hot cell equipment the sample capsule is de- tached from the end of the cable, and a new sample capsule is attached. The new sample capsule is then pulled up into the sampler cavity in readi- ness for another sampling operation. Drive Motor Teleflex Cable 68 Drive Unit Valve Flange Teleflex c0b|e \ Sample Capsule Inner Contui nment Valve Outline ORNL- LR-DWG 75325 Outer Containment Lead Shield Sampler Duct Port Turntable Mount Valve Drive Motor Teleflex Unit Teleflex Drive Valve Drive Motor Turntable. Mount Fig. 4.15. Motor A Inner Containment Operational Sampler Duct Valve <— Quter Containment ™ Maintenance Radioactive Salt Sa.mplexj. Sampler Duct Valve Sampler Duct C 1l 69 The sampler cavity is connected to the hot cell and to each of the four sample points by five separate sample access ducts. Each duct is made of 1 1/2-in. Sch. 40 pipe and contains two 1 1/2-in. double-disc gate valves. The bottom valve is closed only in the event the upper valve is removed for repair or replacement 4.8.13 Coolant Salt Sampling Coolant salt sampling connections for a portable sampling device are provided on the eight coolant loops and the chemical treatment tank. Sam- ples are also drawn from the eight coolant-salt pump bowls and from the make-up salt chemical-treatment tank. A portable; single-sample holding device receives the sample in a manner similar to that of the high level radiocactive salt sampler, transports the sample to the hot cell area, de- posits the sample capsule in the hot cell,;and recelves a fresh capsule from the hot cell. The connection between the portable sampler and cool- ant salt system is of a lock type, has a gas-tight fitting and operates in an inert atmosphere (possibly_radioactive) from full vacuum to 200 psia at 100°F. 4.8.14 Freeze Valves Freeze valves are used to close off gll lines used for transferring salt from one location to another. - These valves are formed in any size of pipe up to 2 in. in diam,by'pinching the pipe to form a rectangular - shaped flow passage. In a 2-in. pipe, the flattened section is about 0.5 in. thick and-2'in;,lcng,'rClbsfiré of the valve is accomplished by freez- ing salt inside the pipe. This is done by directing jets of cold air ‘against the top and bbttdm surfaces of the flattened portion. The air, " supplied by a Rootes-Connersville blower, is controlled to regulate the rate of freezing. Subsequent thawing of the salt plug is achieved in a .féw minfites“by means Qf_a.Calrod'eléétric.heater (3000 watts for a 2-in. line) bent in a.saddleéshapéd series of turns conformal to the flat sec- tion. This heater is easily removed for maintenance. 70. 49 Auxiliary Services and Eguipment - 4.9.1 Introduction Auxiliary services necessary for the MSCR plant are as fiéllcws:1 a. Helium cover gas supply end distribution b. Reagent gas supply and disposal c. Waste gas disposal d. Liquid waste disposal e, Coolant pump lubricating oil system f. Preheating system g. Auxiliary pdwer supply h. Service water systemr i. Control and station air systems j- Cranes and hoists k. Instrumentation and control system 1. Plant utilities 4+.9.2 Helium Cover Gas Supply and Distribution System The helium inert cover gas system serves a number of functions, viz: (a) as a pressurizing gas in various salt containing vessels for the transfer of salt from one place to another; (b) as a carrier gas for the removal of volatile fission products from the recycle gas purge system; (c¢) as an inert atmosphere to protect against contamination of the salt in places such as the UF, system, samplers, melt tank atmospheres, etc.; which are occasionally opened to the atmosphere; (d) as a gas seal and - bearing lubricant in molten salt pumps; (e) to provide the pressures re- quired in the fuel salt and coolant salt circuits to prevent pump cavita- tion, afid to avoid leakage of fuel salt into the coolant salt in case of a leak in the primary heat exchangers; and (f) to pressurize leak-detec- tion devices at various flanged joints and disconnects. | | Since the helium comes into contact with fuel salt or with inter- mediate coolant salt it must be free from oxide-containing impurities such as Hz0, S0, etc; therefore, a purification system is necessary as shown in Fig. 4.16. The raw helium is supplied from a trailer having a capacity He CYL. (12) MANMIFOLD " He TRARLER CONNECTIONS 39,000 3 0-25ctm £V . H, STATION T BF, STATION 3000 REAGENT GAS 200013 o-8scm]__SUELY 0 Fov TO CHEMICAL HF STATION J_] meeament ) f2.0001 BF, TO REACTOR WASTE LIOWHD DISPOSAL L SALT PREP. AND STORAGE AREA 2. FUEL WITHORAWAL AREA 3. 8¢, STRIPPER 4 SALT PUMPS ORNL DWG. 65-7910 VENT AR BURNER ™ KoM = CW. WAKE-UP e . SPENT REAGENT GAS son’ REAGENT GAS 1 m SCRUBBER . DISPOSAL - 0 . FROM REACTOR CELL sumps | SAMLT STORAGE STEAM - AREAS - oewsTER SHPPNG FOOM 20,000-sctm 200 SHIPPY i ‘ M ABSOLUTE AND ROUGHING VENT FULTERS =XPAL oot 1 Cw. . WASTE STORAGE 2-STACK FANS @] T e ; 30-tn. H,0 —~DISCHARGE LOW LEVEL | 20p9ig ' SUMPS ; .‘ 4 c‘ F - B S HIGH LEVEL ~ WASTE CONCENTRATE 6-n.TUBE 3iaTUBE t%-ia. TUBE 160 fi 10N T 0N - MR-COOLED CHARCOAL ADSORBERS DECAY TANK w ) i A LUNES 160 SEMES WATER COOLED CHARCOAL BI“NIMBIIEI ADSORBERS . 8.5 scfm - 6 He RECYCLE 23 pie COMPRESSOR 15%im Fig. 4.16. MSCR Anxiliary Services. WASTE GAS DrSPOSAL 7e of 39,000 std. cublc £t in 30 cylinders at 2400 psi. An emergency supply of 2400 std. cubic £t is supplied from twelve 200 std. cubic £t cylinders. As shown in the flow diagram duplicate lines of helium purification units are provided. Lach has a capacity of 1.5 scfm. 4.9.3 Reagent Gas Supply and Disposal System A reagent mixture consisting of 80% anhydrous hydrogen fluoride and 20% hydrogen is used to remove oxide impurities from reactor and inter- mediate coolant salts. Unreacted reagent gas may contain radioactive material, and thus must be handled by both the gaseous and liquid waste disposal systems. The reagent gases are supplied from high pressure cylinders provided with pressure reducers and flow control instruments so that the flow may be adjusted (Fig. 4.16). The mixed reagent gas is bubbled through molten salt contained in the fuel-salt and the coolant-salt chemical treatment tanks. Unreacted reagent gas passes through a potassium hydroxide (KOH) scrubber to remove HF, and through a hydrogen burner tc remove hydrogen. Spent caustic and condensate from the hydrogen burner may contain radio- active material and thus must be sent to the liquid waste disposal system for concentration and ultimate shipment to a remote disposal area. Non- condensibles from the hydrogen burner are cooled and vented from the sys- tem through the gaseous waste disposal system. 4.9.4 Waste Gas System As mentioned previously, helium 1s passed through the dome above the core to carry away xenon and other volatile fission products. In addi- tion, purge gas is passed down around the shafts of the fuel pumps to sweep away fission products diffusing toward the upper, oil-lubricated bearings. Also, there is helium cover gas in the bowls of the pumps, in the dump tanks, in the fuel handling system, etc., and all of these must be purged to some extent. In order to limit radioactivity in the atmo- sphere surrounding the reactor site, the off-gas from these various sys- tems is passed through charccal beds to trap the fission products until they have decayed, as shown in Fig. 4.16. A few long-lived isotopes, 0 C 73 particularly 85Kr,-decay'but-little in the charcoal beds, and hence con- siderable dilution with air is required to limit the concentration of these in the stack discharge. If the BF; addition system is ever used to effect a "scram" of the reactor, it will be necessary to remove the BF3 from the reactor system before normal reactor operation can be restored. A stripped unit removes BF3 from the recycle helium stream to avoid saturating the off-gas ad- .. sorber as shown in the flow diagram (Fig. 4.16). 4.9.5 TLiquid Waste Disposal System Liquid wastes originate in the drains and sumps of the various equip- ment cells. Typical sources are hot sinks in the analytical laboratory, cell drains in the shipping area where the shipping cask will be flushed with decontaminants, the contaminated equipment storage and decontaminasi.c tion cell, and spent caustic solutions from the reagent-gas disposal sys- tem. The liquid waste system shown in Fig. 4.16 provides for holdup, con- centration, and storage of active wastes on-site. High level wastes are concentrated by evaporation and stored in underground tanks. Low-level wastes may be held in underground tanks or in a low-level activity pond until they have decayed sufficiently for discharge into the river. The capacity of this pond is approximately 25,000 cubic ft (nearly twice the volume of the contaminated equipment storage cell). Intermediate wastes are-sent to ten 10,000-gallon stainless steel retention tanks, which are ‘used: for:temporary storage until it is possible to send the waste to an evaporator for concentration. High level wastes from the plant and from the evaporator are sent to a 10 OOO-gallon-waste—storage tank which provides semi-permanent storage of these wastes, which will ultlmately be transferred to shlelded shipplng containers for shlpment to a permanent disposal area. Spent caustic wastes from the reagent gas disposal system are sent to the 1ntermed1ate-level storage tank for further concentratlon and ultimate dlsposal off- 81te | Pumps (mechanlcal, air Jet, or steam jet) areiprovided at every sump, at the pond, at the 10,000-gallon tanks, at the evaporator tank, and at the 1000-gallon tanks to transfer the liquids. 7, | The waste evaporator is capable of'evaporating2QO_gallpns of water per hour using the plant'heating steem. The-condenser‘diaihs either to the retention tank or to the low level fiaste pond,fidepending on the ac- tivity. A vent on the shell side allows noncondensibles to be vented to the charcoal beds. 4.9.6 Coolant Pump ILubricating O0il Systems These systems supply oil to the.bearings of eight pumps in the fuel circuits and to eight pumps in the intermediate coolent loops. The two groups of eight pumps each are supplied by two independent systems. The lubricant is circulated under oxygén-free conditions, 120 gpm to each pump group. The systems asre designed to 120°F supply with an. expected 0il temperature rise of 20°F. 4.9.7 Preheatigg;System Prior to the admission of salt, all salt-containing piping and equip- ment must be preheated to 900°F . This is accomplished by the use of - several types of resistance heating. In sizes up to about 2 in., piping is heated by passing an alternating current through the piping'itself. All pipes heated this way are electrically insulated. Larger diameter salt-containing pipes are heated by hinged resistance heaters (2000 w/ft) surrounding the pipe. For large pipes (above 10 in.) a three-section heater assembly is used. Piping fittings are covered by prefabricated heating units. Heaters for pumps are field-fabricated. The reactor vessel and primary heat exchangers are heated by tubular . resistance heaters attached to the outer surfaces by means of clips on -welded studs. 4.9.8 Auxiliary Power . Auxiliary power is supplied at 4160 volts by e pair of auxiliary power transformers, type OA/FA, 24-4.16 kv with. a maximum fan-cooled rating of 25 Mva. A reserve transformer, fed from the switchyard bus, serves as a standby unit. The loads fed at 4160 volts include all motors rated at 150 horsepower or more, and the suxiliary power transformers O % 3 “ v, 75 required for station lighting, 480 volt auxiliaries, and the fuel melting and preheating systemn. _ Smaller auxiliaries, of 30 horsepower or less, and those whose con- tinuous operation is not considered vital to station operation, are fed from motor control centers in the vicinity of the load. Two 15 kva 480- 120 volt transformers supply a 115 volt a-c control and instrumentation bus for each unit. An emergency supply to this bus is provided from a 15 kva inverter motor-generator set which is driven by the unit 250-volt battery. 4.9.9 BService Water System The service water system supplies river water for cooling purposes throughout the plant, includihg the reactor auxiliaries and the turbine plant components.. | Water is supplied by three 15,000 gpm, half-capacity, vertical cen- trifugal pumps. Each pump is driven by a 1250-horsepower motor, and is located in the circulating water intake structure. During normal opera- tion two‘pumps-supply the system with water at 100 psig, with the third pump employed as a standby. 4.9.10 Control and Station Air Systems The compressed air system for the plant consists of separate, inter- connected air supplies for the station and for control purposes. The station air system supplies hose w_r_a.ives for operating and meintenance re- quirements throughout therstationQ Control éir is used'pfimarily for in- strument transmitters and'airaOperated'valves.- The two*airgsystems are cfosé.cofinected_so that ébmpressed.air may be supplied to the control air system in event of a compressor failure. | | | - The control air system of the plant supplies air-operated control devices at & header pressure of 115 psia'énd iSfreduced to 55 psia and 45 fisia for supply to various drive units and instrument transmitters. 4.9.11 Cranes and Hoists A single traveling bridge crane serves the reactor and steam turbine buildings. Its lifting capacity is based on handling the rotor of the 76 low pressure steam turbine-generator, which weighs about 150 tons. The bridge span is 130 ft, and the crane'lift is sufficient to reach the.low-_» est portion of the building. All heavy equipment coming into the building by rail may be handled by the crane. Its capacity is sufficient to allow removal of all components>within the reactor primary shield eicept the reactor. _4.9.12 Instrumentation and Control The requirements for instrumentation and control of the turbine sys- tems are similar to those of the turbine system in a conventional fossil- fuel power plant. | Instrumentation for the reactor system.monitofs the reactor neutron flux, primary system pressures, temperatures, levels, and flow rafiés, and provides control and alarm signals to actuate the appropriate device or to call for operator action when changes occur in the measured quantities, through either changes in load or malfunction of system components. Control and instrumentatibn panels are located in the control room, for convenience of reading, recording and operating the most important quantities and components. Other auxiliary control panels or isolated instruments may be located at appropriste places in the plant; area radia- tion monitors, alarm or warning signals, hydrogen and seal 0il controls for the generators, etc. 4.9.13 Plant Utilities The plant utilities include those systems that are provided for moni- - toring plant equipment, dispoéing of nonradioactive wastes, safety of personnel, protection of equipment and for heating, ventilating and air- conditioning the plant buildings. Theée systems do not differ appreciably from those provided for conventional plants. ' ¢ " I “w w) 77 4.10 MSCR Design Specifications .The significant specifications for the MSCR equipment and materials are listed in Tables 4.2 through 4.6. The vessels are described in Table 4.2, heat-transfer equipment in Table 4.3, pumps and circulators in Table 4.4, miscellaneous equipment in Table 4.5, and materials in Table 4.6. Table 4.2, INOR-8 Salt-Containing Vessels Heater Weight Number Volume of Wall Design Design of Each gnit Diar:gter Height Thickness (ggggci;{t) Temp . Pressure l]gegggs - . o ) Units £t in, o F psl 1b tare 1. Reactor Vessel 1 630 fuel 2kl 240 1.7 1100 inlet 100 in 1,100,000 1300 outlet 50 out 2. Fuel Salt Drain 5l 50 35 96 1/2 15 1300 100 ‘ . 3. Fertile Salt Melt 1 150 42 bk 1/4 150 1200 50 Tank 4, Fertile Salt Chemical 1 100 36 168 5/8 15 1300 100 - 25,000 g Treatment Tank : 5. Fertile Salt Storage 1 100 36 168 3/8 12 1200 100 23,000 g 6. Fertile Salt Small 1 2 12 32 3/16 1 1300 50. 500 g Addition Metering : ' Tank _ Q 7. Fertile Salt large 1 50 36 90 1/8 10 1000 100 o Addition Metering Tank | 8. TFuel Salt Small 1 2 4 300 0.4 12 1300 150 Withdrawal Metering Tank 9. Fuel Salt Reactivity 2 50 36 96 1/2 10 1300 100 Control Drain Tanks B 10. Fuel Salt Decay 5 35 30 90 1/4 10 1300 25 7360 g Storage Tanks "900 ¢ 11. Fuel Salt Withdrawal 1 10 30 27 1/4 3 1300 25. 23%0 g Transfer Tank ‘ 290 ¢t 12. Flush Selt Storage 5 650 96 156 1/2 150 1000 : Tanks . 13. Coolant Salt Melt 1 150 42 1y 1M 150 1200 50 Tenk 14. Coolant Salt Chemical 1 50 42 60 1/2 8 1300 100 11,500 g Treatment Tank ) 15. Coolant Salt Drain 2 750 43.7 864 1/4 200 1100 80 100,000 g Tanks 10,000 t » o " ) Table 4.3. Heat Transfer Equipment Primary Heat Stean Steam Loeffler e et Exchangers Superheaters Reheaters Boilers c . ondensers AEC Account Nmnber 221.314 222.32 222.322 222.31 223.312 Design data " Number of units 8 16 8 4 1 Unit heat rate, Btu/nr 1,066 x 10° 467 X 108 132 x 10° : 85 x 106 - Geometry Shell g U-Tube Shell & U-Tube Shell & U-Tube Cyl. Drums Shell g U-Tube Number of tubes 2,025 785 766 None 630 Active area, ft? 6,643 8,905 3,543 1,036 Active length, ft 25 57.8 23.6 100 10 Length of longest *ube, 28.8 61.8 £ Length of shortest tube, 23.7 53.8 £t , " Tube OD, in. - - 0.5 0.75 0.75 0.625 Tube-wall thickness, in. . 0.035 0.083 0.065 0.065 Iattice pitch, in. ' 0.625 .00 1.5 Tube material : INCR-8 INOR-8 INCR-8 Admiralty Shell material INOR-8 INOR-8 INCR-8 Carbon steel Carbon steel Shell ID, in.. . - 4375 -31.5 31 72 26 3 - Shell thickness, in. - 1.5 0.5 0.5 7 0.25 O IMID, °F 173.7 174.8 187.5 Shell weight, 1b 16,000 11,000 4,500 Tubing weight, 1b 10,000 25,000 9,620 3,800 Design pressure Heat transfer coefficient, 924 300 119 Btu ' hr-£t2:°F Shell-side conditions Fluid ‘ Fuel salt no. 2 Int. cool. salt Int. cool. salt Stean Steam Inlet temperature, °F 1,100 1,100 1,100 240 1,000 Outlet temperature, °F 1,300 950 950 120 636 Flow rate, ft>/sec 20.25 13.7 3.84 80,000 3.7 x 10° Pressure, psig 200 300 300 10 2,460 Pressure drop, psi 80 15 40 Volume, ft> 61.6 150 64 Tube-side conditions Fluid Int. cool. szlt Steam Steam Water ‘Water Inlet temperature, °F 950 670 635 80 545 Outlet temperature, °F 1,100 1,000 1,000 150 636 Flowrate, 1b/hr 31 1.28 x 10° 0.7 x 108 1.2 x 108 5.2 x 10° Prezsure, psig 350 2,490 440 110 2,450 Pressure drop, psi 84 25 20 50 Table 4.4. Pumps and Circulators pump housing flange face (Approx.) Pump motor rating Pump motor speéd (synchvonous) Pump motor type | ' 1600 hp, 4160 volt 900 rpm Totally enclosed, 3 phase water cooled MSCR Fuel Pump CoolgigRPump Cirensators Number of units 8 8 4 Type Centrifugal Centrifugal Axial Fluid pumped MSCR-2 fuel salt Constant salt Steam Service temperature 1300°F 1100°F 1000°F Fluid density 190 1b/ft 120 1b/ft> Fluid flow per pump 9075 gpm 13,900 5.3 x.10° 1b/hr Suction pressure (17 ft NPSH) 22.5 psia 130 psia 2430 psig ‘ Discharge pressure {150 ft developed 220 psia 350 psia 2490 psig head) - Impeller OD 25 in. 25 in. Suction ID 14 in. 14 in. Discharge ID 12 in. 12 in. Over-all pump OD (Approx;) .. 50 in. 50 in. Over-all pump height-suction opening to 50 in. 50 in. 2000 hp, 4160 volt 900 rpm Totally enclosed, 3 phase water cooled -0 O O 9 81" Table 4.5. Miscellaneous Equipment Thermal shield Dimensions, ft Material Shield Headers 1.5 X 24 x 24 2 in. plate carbon steel 20 in. X 4 in. carbon steel (1/2 in. wall thickness) Weight, 1b gross (water filled) 250,000 Inlet/outlet pipe 8 in. Sch. 20 carbon steel Heat removal rate, 10° Btu/hr 18 Water flow, gpm Water temperature/pressure, °F/psig Shield cadling system Heat load, Mwt Demineralized water circuit °F °F Maximum temperature, Minimum temperature, Flow rate, gpm | Service water conditions Maximum temperature, °F Minimum temperature, °F Flow rate, gpm Cell-air coolingrsystem Heat load, Mwt Demineralized water circuit Meximum temperature, °F Minimum temperature,.fipm' Flow rate, gpm Service water éqnditibns ‘Maximum temperature, Minimum temperature, Flow rate; gpm °F °F 2000 90-110/15 37.5 180 125 4650 125 75 5120 110 95 2275 95 75 1700 R - 82 Table 4.5 (continued) High level radicactive salt sampler Sampler cavity linner containment vessel Dimensions Material and thickness Design temperature/pressure Sampler cavity shielding Material; thickness Approximate weight Sampler cavity outer containment vessel Dimensions Material and thickness Design temperature/pressure Sampler ducts Design pressure Design temperature Size Material Sampler duct valves Design pressure/temperature Number required Type Cable drive unit Service environment 5 ft 4 in. ID X 6 ft high 304 stainless steel, 1/8 in. 125°F/+15 psig Lead; 1 £t thick all around sides and top 57 tons 7 £t 8 in. ID X 8 ft high Carbon steel, 1/4 in. thick 125°F/15 psig 100 psig 200°F 1 1/2 in. Sch. 40 pipe Inconel 100 psig/125°F 10 : 1 1/2 in. Vulcan bellows stem double-disc motor-operated gate valve (supplied by Hoke Valve Co. ) Drive unit will operate in an inert but radiocactive atmosphere at normal temperature. Radia- tion level will be high only during sampling operation. The cable must be capable of operation at temperatures in the range 100 to 1300°F. O is i “m ) w 83 Table 4.5 (continued) High level radioactive salt sampler (continued) Cable drive unit (continued) Cable size UF,; addition facility Containment vessel Dimensions, ft Design pressure, psig Material Pellet bin and dispenser Volume, £t Metering chamber Configuration Tube length, ft Tube material Valves Size (nominal) in. Description UF,; addition pipe Dissolver tube Configuration ~ Size, nominal pipe size (in.) Material Length, ft Other description - Shielding 1/8 in. diam; length sufficient to reach from the sample cavity to the farthest sample point. 3 x 10 +15 304 stainless steel Coiled tube, 1 in. pipe 14 Stainless steel 1 Bellows-sealed, 150-1b design, gas-tight, soft-seated 1 in. Sch. 40 stainless steel pipe - Perforated, coiled pipe L INOR-8 4 End blanked off; perforation ~ diameter, 1/8 in. 6 in. of lead 8 Table 4.5 (continued) Reagent gas disposal system data - Hydrogen supply KOHvsupply tank Cylinder station capacity, 3000 standard cubic ft Flow rate, scfm -2 Hydrogen fluoride supply Cylinder station capacity, 12,000 standard cubic ft - Flow rate, scfm 8 KOH scrubber for HF disposal HF flow rate (max), scfm 8 Cooling requirement, Btu/hr 118,000 Cooling water flow rate, gpm 10 KOH solution feed rate, 1 liters/min (max) Flow rate of unreacted 2 hydrogen, scfm Length and diameter, ft 8 and 2.8 Material of construction Monel Wall thickness, in. 1/4 Dimensions 4 £t diam, 8 ft high Material Carbon steel Chemical feed pump 0-1 liter/min Mixer 2 hp Hydrogen burner Design flow rate, 1b-moles Hp 1/3 per hour Heat load, B/hr 41,000 Cooling water flow, gpm 4 Design air flow rate, cfm 10 (max) Dimensions | Material 1 ft diam x 4 ft high Carbon steel i k1] » 85 .Table 4.5 (continued) Reagent gas disposal system data (continued) Hydrogen burner condenser Heat load, B/hr Cooling water flow, gpm Exit gas flow rate (nitrogen plus unburned oxygen), scfm Design temperature of exit gas, °F | Surface area of tubing, 2 Shell material Tubing material Configuration Condensate rate, 1b Hp0 per hr Cover gas purification system Helium dryer | Dessicant used Amount, 1b Length/diameter, ft/in. Pressure rating, psig Container material Helium heater (electric) CF 'Helium.flow'rate,'scflm - Design pressure, psig . Heater rating (electrié), kw Oxygen removal unit Design flow rate, scfm 'Container'size,'length/diam— eter, in./in. (overall) Active ingredient Design temperature gas exit, o : T : 20,000 100 100 Carbon steel Admiralty Shell-and-tube, single pass, straight tube 6 Molecular sieve 10 2.5/4 300 Carbon steel 1200 1.5 300 -1 1.5 26/6 Sch. 40 Titanium sponge 86 Teble 4.5 (continued) Cover gas purification,syStem:' (continued) Oxygen removal unit (continued) Design pressure/temperature, psig/°F Helium cooler 4 in. finned tube with flanges Inlet/outlet temperature, °F Helium flow rate, scfm Treater helium surge tank Tank volume, ft> ' Design pressure/temperature psig/°F | | Material Lube o0il system Number of systems Design lube oil flow, gpm Design lube oil temperature, °F (in/out) Design cooling water tempera- ture, °F (in/out) Design cooling water pressure, psig Design cooling water flow, gpm Number of lube oil pumps Type “"* Lube oil pumps head, ft Flow, gpm Motor, hp Reservoir, number 2 ID, £t (cylindrical) Height, ft Capacity, gal. Wall thickness, in. 300/1200 4 £t long 1200/200 1.5 60 300/100 Carbon steel 120 140/120 75/100 25 45 Rot. 150 300 20 600 0.25 O " Hy " Table 4.5 (continued) Lube 0il system (continued) Reservoir, number 2 (continued) Material Weight (full), 1b Cooler, number Type Tube material Shell material Overall heat transfer coefficient, Btu/hr-ft2.°F Tube surface area, £t Tube size, OD, in. Tube wall thickness, in. Tube pitch (triangular) Shell diameter, ID, in. Shell length, ft Tube design pressure, psig 4 Tube design temperature, °F Shell design pressure, psig Carbon steel 3000 2 2 pass shell, 4 pass tube Inconel Carbon steel 20 735 5/8 0.125 0.938 18 34 25 100 45 88 Table 4.6. Material Spécifiéations Properties Assumed for MSCR Graphite Items and Units MSCR MSRE¥ Density, g/cc S 1.9 1.9 Thermal conductivity, Btu/hr* | ft°°F ' o At 68°F and with grain 80 At 68°F and across grain 45 At 1200°F, isotropic, after 15 . irradiation . ‘ Coefficient of thermal ex- pansion, across grain, per °F At 68°F | 1.7 x 107° At 1200°F, after irradia- 3 x 10~ tion Maximum allowable strain, 0.001 in./in. , Porosity _ _ Accessible to salt at 150 0.001 . 0.005 psi | ' Accessible to gas 0.01 Poisson's ratio 0.4 Young's modulus With grain ~ 3% 108 . Across grain 1.25 x 106 1.5 x 106 Helium permeability at 30°C, | . f ‘cm? /sec | Diffusivity of xenon at 1200°F, 10-6 cm? /sec Specific heat Btu/lb-°F 0.33 *MSRE values are given for comparison and were mostly taken from reference (12). O " Y] # Table 4.6 (continued) 89 Y t—— Assumed Properties of MSCR Fuel-Salt Mixture at 1200°F Mixture No. Composition Mole % LiF-BeF,-ThF, Wt % LiF-BeF,-ThF, Liquidus temperature, °F Molecular weight Density, 1b/ft> Viscosity, 1b/ft-hr Thermal conductivity, Btu/hr«ft-°F Heat capacity, Btu/lb:°F Assumed Properties of MSCR Intermediate 1 71-16-13 29-11-60 941 66.03 215.6 24.2 2.67 0.318 Coolant Salt Mixture at 1062 °F* Composition Mole % LiF-BeFa Wt % LiF-BeF, Liquidus temperature, Molecular weight Density, 1b/ft> Viscosity, 1b/ftshr Thermal conductivity, Btu/hr-ft-°F OF‘ ~ _Heat capacity, Btu/1lb:°F *Reference (12)*of the Bibliography. Properties of INOR-8% Chemical composition Nickel,_min Molybdenum 2 3 68-23-9 66-29-5 32-19-49 38-29-33 887 860 56.2 46.2 190.1 163.0 21.0 18.9 2.91 3.10 0.383 0.449 6634 5248 851 33.14 120.5 20.0 3.5 0.526 Wt % 66~71 (balance) 15-18 *Reference (12) of the Bibliography. 90 Table 4.6 (continued) Properties of INOR-8 (continued) = Wt % (continued) Chemical composition (continued) a Chromium o 6—8 Iron, max - 5 Manganese, max i 1 Silicon, max - o | Carbon | ~ 0.04-0.08 Miscellaneous, max _ 2 Physical Properties at 1200°F Density, g/cc 8.79 Melting point, °F | 24702555 Thermal conductivity, Btu/hr-ft-°F 11.7 at 1200°F ' = Young's modulus, psi at 1300°F 25 x 108 Specific heat, Btu/1b-°F at 1200°F = 0.1385 Coefficient of thermal expansion, 7.8 x 1076 1/°F at 1200°F | Meximum allowable stress, psi at 6000 1200°F Maximum allowable stress, psi at 3500 1300°F ' O [ ] [Y] " 91 5. TFUEL PROCESSING 5.1 Reprocessing System Fluorination of spent fuel from the MSCR to reoover isotopes of ura- nium, followed by discard of the carrier salt (containing LiF, BeF2, ThFy, and fission product fluorides), was selected as the method of processing for several reasons: (a) The fluorination process is well adapted for future integration with the reactor plant (sharing shielding and mainte- nance facilities and personnel) with appreciable potential reductions in fuel cycle cost (Sec. 6.9); (b) thetneoeSSity for holding the spent fuel for decay of badioactivity:is'eliminated; (¢) no development is required; the plant may be designed and costed on the basis of current technology. In order to use the next.fiost applicable process, Thorex, it would be necessary to develop special head-end and tail-end steps for converting spent fuel from the fluoride to the nitrate and back again. Also it would be necessary to hold the spent fuel prior to processing for not less than 120 days (to average the equivalent of 90 days of cooling). Further, there is some‘uncertainty concefining'the effect of fluoride ion on the chemistry of the aqueofis separationé; On ‘the other hand, the thorium could be recov- ered, and perhaps also the carrler salt could be recovered free from con- tamination with rare earth 1sotopes.: It would, however, be contaminated with the isotopes Cd4d, Sr;oAg; Cs,zse, Ba, and Te. 'The processing-costs for the'MSCR;fuel have been estimated for both processes, i'5.2!_r1uor1ae—leatility Central Plant 3 - A central plant oapable of proce531ng about 30 £t of fuel salt per ' fl'day (l tonne Th/day) was selected for oostlng. Thls plant is capable of - ,servxclng about 20 reference deszgn MSCR's having a total capablllty of 20,000 Mwe. | | S Component de31gn, plant layout, and associated costs fbr the plant 'descrlbed_hereln_were adapted from a design .and cost study of an on-site ‘'plant prepared by Carter, Milford, and Stockdale in a prior study (21). 92 Due allowance for fuel transport to a central location was made; together with other adjustments for difference in capacities, elimination of prot- actinium recycle, etc, 5.2.1 Process Desi&a The steps of the proposed fluorination process are indicated in Fig. 5.1, Spent fuel is transferred from the shipping containers to pre- fluorination storage by applying gas pressure or by siphoning. = The fuei is then introduced batch-wise into the fluorinators where it is treated with elemental fluorine, possibly diluted with some inert gas, at about 1000°F, The effluent UF6 is absorbed in beds of NaF at 200°F,,is«iater desorbed at 700°F and collected in cold traps at -45°F, Periodically the cold traps are warmed, and the decontaminated UF6 is ‘collected in cylinders, The fluoride volatility process does not provide for recovering thorium or any of the componénts of the carrier salt. Conseqfiently,after fluorination the LiF-BeF —ThFu-FP melt is drained into interim waste stor- 2 age, and later transferred to permanent waste storage such as,.for-example,' - in a salt mine., The interim storage period has been taken to be 1100 days, a value corresponding to the most favorable economic balance between on- site and permanent storage charges for this particular process, It will be observed that 233pa is not recovered in this process. An analysis of Pa recovery versus discard disclosed that additional process equipment and building space requirements made the recovery of Pa. uneconomical., , A major problem in the desigfi of vessels which contain irradiated salt is that of heat removal, Volumetric heat release rates are high, and the temperature of the heat source is considerably greater than that of con- ventional heat sinks (such as cooling water from rivers or wells), In the design evolved, heat is transferred acroés an air gap into water, The principal heat transfer mechanism is radiation; convection accounts for perhapsrs'to'lo per cent of the transfer. This arrangement, in addition to controlling the heat transfer rate at tolerable levels, provides'iso- lation of the coolant from the moltep_salt so that a leak of either stream through its containment wall does not contaminate the other stream. — - % Al ] o . - & & .1’ ) ORNL-LR No. 74670 ;’ \ Cooling Water tock Dtschurge To River Q - ' erSupply 1 Air - § Litntoke 2 r fl L I Product Recov sl | | ! i l ' l i rrynl —— . —— —— l60 PREFLUORINAT]ON L ' STORAGE TANKSI e i e e o e o i o ol €6 Filter e el it o i e, e | i | I | I | ! I I i i I L _—““—_——-—“-——J L Li;';u'gl Fp Recycle - . : : t l . I Collection NaF " UF, Recycle 1o ‘ : 9 4 ——) | Cylinder Bed Sed Cooling Wotet —— e e e Reactor - ‘ j‘ Gl Intake Fljom River L_ A l LIF- BeE, - ThE, + Fission Products L I L E e — - —— === e T T T — | :.. | l ‘ | — e | Water Discharge — e E—— == == e _ To River “"':‘ - = = =T = = = = T =T ! Fig. 5.1. MSCR Fluoride Volatility Fuel Processing Plant. 9% The fluorinators are cooled by circulating air through the cell. 1In the case of the prefluorination storage tanks, radiation from storage vessels to a concentric tank in a water bath provides sufficient cooling for 5-day-old fuel salt, Wherever possible, the equipment was patterned after that used in the ORNL Volatility Pilot Plant described by Milford (66), Carr (18), Cathers et al. (22, 23), : 5.2,2 Shipping Shipment of irradiated fuel is made in a lead shielded carrier shown in Fig. 5.2. The cask is equipped with a water-cooling system which is able to absorb decay heat radiated from the salt container and to disSipate this heat to the atmosphere via finned tube exchangers fastened to the out- side of the cask., Heat transfer may be either by boiling the water in fhe inside jacket followed by condensation in the outside exchanger, or by natural convection. | | The shipping container (also used for waste disposal), is designed to hold ~10 £t3 salt which is about six days accumulation at the reactor dis- charge rate of 1,67 ft3/day. For this design it has been assumed that the processing plant is located 500 miles from the reactor site and that ship- ment will be made by rail. The round trip, including filling and emptying, is anticipated to take 10 days. The average age of spent fuel at shipment is approximately 5 days. 5.2,3 Prefluorination Storage Tanks 3 tanks receive up to 120 days supply of One hundred and sixty 30 ft spent fuel from each of a number of reactor sites., When a 120-day batch is completed, the tanks are removed from their cooling jackets to .the transfer area where the material is transferred to 6-ft3 metering transfer tanks, From these tanks the spent fuel can be transferred by gravity or inert gas pressure to the fluorinators. The cooling jackets for the prefluorination storage tanks are stéin- less steel tanks 2.2 ft in diameter by 12 ft high immersed in a water bath. Cooling is achieved principally by radiation from the storage tank to the <;>_ ¥ ¢ 95 ORNL-DWG. 65-7911 Lead Shielding (14.5 thick) / / 7777 o4 Finned Tube Condenser or )(//r—.Heat Exchanger / '/ 7/ Salt Container 71777 Fig. 5.2. \\\u;Jacket for H;0 Circulation Irradiated Fuel Shipping Cask. 96 jacket. The transfer tanks are'equipped with jackets cooled by circulating water., Electric heaters provide preheating prior to transfer operations, should heating be required., 5.2.4 Fluorinator The design shown in Fig. 5.3 has been successfully operated in the ORNL Fluoride Volitility Pilot Plant (18)., Surmounting the fluorination chamber is a de-entrainment section. The lower chamber is surrounded by an electrically-heated furnace while the upber is heated with electrical strip heaters, Five units are required, each having a capacity of 6 £t3, The corrosion rate is about 1 mil per hour of fluorination time; hence the fluorinator must be inexpensive and accessible for frequent replacement, It was designed to dissipate decay heat and heat of reaction to the atmos- phere in the cell through a wall 1/2 in. thick at a temperature of SO00°F. The preferred materials of construction are either INOR-8 or Alloy 79-4 (70 per cent Ni, 4 per cent Mo, 17 per cent Fe). L-Nickel has been used for fluorinator construction but is susceptible to intergranular attack. Spent fuel is fluorinated batchwise at about 1000°F., It takes about 6 hours to volatilize the uranium (99.9+%) from a 6 ft3 batch, In current practice the attack of fluorine on the vessel is severe, The high rate of corrosion is believed to result from the combined action of liquid salt and gasedus fluorine phases, However, several lines of improvement are under investigation., These include fhe use of the "frozen wall" fluori- nator (35) wherein a layer of solid salt is maintained on the vessel wall by proper control of the cooling, and this layer protects the wall, Another approach consists of spraying the molten salt into a relatively cool atmosphere of fluorine. Uranium hexafluoride is formed in and rapidly removed from the microdroplets which then cool and freeze before they strike the wall. Not only are the wall temperatures lower, but there is no liquid phase in contact with the wall, | | The fluorides of some fission products, notably Mo, Zr, Nb, Cs, Ru, and Te are volatile and accompany the UF These are separated from the 60 uranium in the CRP trap and NaF absorber described below. It is not ex- pected that the fluoride of Pa will be volatile under conditions specified, o n & ' ~134-in. NICKEL 20l 1 & i 'SAMPLE LINE 1|||;g;| (Tin NPS SCH. 40) TUBULAR ELECTRIC L HEATERS - CORROSION-SPECIMEN NOZZLE(6) ~ N} FURNACE LINER —] | TREPANNED SECTION (4-in. OD) DRAFT TUBE —| / WASTE-SALT OUTLET —MATERIAL ORNL-LR-DWG 39150-R2 o7 PRODUCT OUTLET TO MOVABLE BED ABSORBER [====28 = PRODUCT OQUTLET | * (ALTERNATE) SOLIDS SETTLING GHAMBER VESSEL SUPPORT CAPPED NOZZLE i o] -_?_\ q /4 q 1 Al 1 1 g g LU % o g_‘u :; — FLUO QOOOOTROOOO0O0ODOOOCOO] D000 U0 T OO0 0N 00000000 " FURNACE THERMOCOUPLE WELL (16-in.0D) RINE INLET * (4-in. O 0 5 10 INCHES ' Fig. 5.3. Fluorinator. __ FLUORINATION CHAMBER \__———-TREPANNED S)ECTI_ON 15 20 98 Protactinium will remain in the barren salt and be lost to waste. The loss amounts to about 10 g/day in the reference design reacfor, and studies have shown that it is not economical to recycle the barren salt to the fluori- nators after holding it to allow the Pa to decay. The cost increase of additional fluorinator capacity and interim storage‘vessels more than off=- ‘sets the value of the 233y pecovered. 5.2,5 CRP Trap and NaF Absorbers After leaving the fluorinator, UF. and accompanying fission product fluorides pass into a two-zoned NaF ab:orption system, The first, called the complexible radioactive products (CRP) trap is operated at about 400°C and removes fluorides of chromium5\zirconium,,niobium, cesium, strontium, and rare earths, as well as entrained salt partiéles° Uranium hexafluoride is not absorbed here, but is absorbed ifi-the second zone operated at 100°C, along with the fluorides of molybdenum and ruthenium, and traces of others. Some ruthenium carries through into the fluorine recirculation system. Uranium is recovered from the beds by desorption at 400°C, It is collected in cold traps described below. o The stationary bed absorber, shown in Fig. 5.4, contains just over one cubic foot of NaF, Six unites are required. Each is mounted in a lightweight, low-heat capacity electric furnace which opens on-hinges for removal of the absorber. A cooling-heating tube 2-1/2 inches OD carrying coolant and containing electric heaters extends through the center of the bed., An interior cylindrical baffle forces the process stream to follow a U-shaped path through the bed, ' Design limitations arise in the rate at which the bed'temperature can be cycled and the bed thickness, The granular bed is a rather effective insulator and must be made in thin sections to facilitate heating and cool- ing. The absorbers therefore have large length-diameter ratios, When the bed becomes saturated with fission products, the absorber is removed from the furnace, emptied, and recharged remotely in a 4-5 day cycle, 99 ORNL-LR-DWG 39257 COOLING AIR INLET INLET;L FOUTLET ‘—T""'fi_"l [% \ Y2-in. NPS — | 3 SCHED 40 || | ELECTRICAL ROD- TYPE HEATERS, 4000 -w TOTAL Yg X Yg-in. DIA %% NaF PELLETS/~ i 6-in. NPS SCHED 40— 7} 4Y5-in. OD x /g -in. WALL il 0 1l SO | 2Y,-in. OD x 1/g-in. WALL '/4 in. PLATE/ - 1 - 1 { —14 \ N N H SE=S - SiL-0-CEL INSULATING POWDER INCONEL MATERIAL : 24 in. THERMOCOUPLE WELLS Fig. 5.4. Sodium Fluoride Absorber. 100 $.2,6 M These are similar to those used in the ORNL Fluoride Volatility Pilot Plant as shown in Fig., 5.5. Two traps are mounted in series., The first is operated at about -u40°C and the second at =60°C, Adequate surface for rapid transfer of heat and collection of solid UF6 must be provided. The components must have small thermal inertia so that the temperature may be changed quickly. During defrosting, the traps are heated to 90°C at 46 psia to allow UF_ to melt and drain to collection 6 cylinders, 5.2.,7 Reduction Reactor The reduction of UF6 to UFl+ is accomplished in a reactor patterned after that described by Murray (70) and consists of a 4-in. diam by 10-ft high column, Upranium hexafluoride and flourine are mixed with excess H2 in a nozzle at the top of the reaction chamber. The uranium is reduced in the BQ- _ in molten salt of suitable composition. Gaseous materials are discharged F, flame, and falls to the bottom of the chamber where it is collected through a filter. The reactor has a capacity of 10-15 kg of UF6 per hour, Losses are very low and typically are less than 0.1l per cent. 5.2.8 Transfer Tanks Stripped fuel is drained from the fluorinators into transfer tanks (two each) from whence it is distributed to interim waste storage tanks, The transfer tanks have capacities of 60 £t3, Decay heat is radiated from the surface of the tank through a 1/2 in. air gap to a water-cooled jacket. While being held in the transfer tanks, the salt is treated with He or other inert gas to remove tfaces of F2 of HF that would increase corrosion in the waste storage tanks, 5.2,9 Waste Storage‘Tanks Waste salt is stored in stainless steel shipping cylinders 2 ft in diam by 8 ft long. These are placed at the bottom of steel thimbles 2,75 ft in diam and fifteen feet long which dip into a water-filled canal. Heat is dissipated by radiation and convection across the 4-inch 101 ORNL-LR-DWG 19091 OUTLET WELL HEATER o < w - W wd - 3 FILTER CARTRIDGES REFRIGERANT TUBES (4) 5-in. SPS COPPER PIPE 24 12 SCALE IN INCHES 12 CALROD HEATERS (6) INLET END HEATER " PRIMARY COLD TRAP Primary Cold Traps. 'Fi'g. 5.5, 102 air gap, through the thimble wall and into the water, After cooling, the tanks are shipped to a salt mine for permanent storage. 5.2,10 Freeze Valves Conventional valves cannot be used with molten salts. Flow stoppage is achieved by freezing a plug of salt in a section of a line with a jet of cooling air. Electric heaters are used to thaw the plug when flow is desired, A freeze valve for the MSRE is shown in Fig. 5.6. 5:,2,11 SamBlers The apparatus pictured in Fig, 5.7 is being tested for use_wifh the - MSRE (12), Essential features are the hoist and capsule for removing the sample from the vessel; a lead-shielded cubicle with manipulator, heating . elements and service piping, and a transport cask for removing the sample from the process area. The sampling cubicle is mounted on the cell biolo- gical shield in an accessible area. 5.2,12 Biological Shield Calculations were made using the Phoebe program for the IBM 704 com- puter, In the study by Carter, Milford, and Stockdale (21), on which the present estimate is based, spent fuel was brought to the processing plant immediately after removal from the reactor, and the shield was accordingly made quite thick, In the present instance, the fuel is cooled at the pro- cessing plant for not less than 90 days so that the shielding requirements are not as extreme, However, in order that the central plant héve more general utility, no reduction in shield thickness was made. 5.2.13 Process Equipment Layout Process equipment is laid out according to the major process opera- tions: prefluorination storage, fluorination, transfer, NaF absorption, cold traps and product collection, UF, =+ UFu reduction, and interim waste 6 storage. Equipment is grouped in cells according to activity level and in an arrangement that minimizes distances between vessels. Three transfers i;;- of molten salt are required in the processing sequence. - - H bt _Fhoto 36713 103 Freeze Valve. Fig. 5.6. 104 SAMPLE TRANSPORT CONTAINER EXHAUST HOOD ORNL-LR-DWG 59206R TRANSPORT CASK AREA 4 # _SAMPLE TRANSPORT CONTAINER /"REMOVAL VALVE - PERISCOPEC EXHAUST CABLE DRIVE - : MECHANISM A A, A ISR HELIUM : ‘o Q_O RN . PURGE ~—LATCH O CAPSULE R0 o URANIUM ' - VACUUM SHIELDING AREA 3 '"G\n D4 PUMP OPERATIONAL | VALVE _ LEAD — SHIELDING MAINTENANCE VALVE - [v}f—r— AREA 2B .'A_::: A Y I .';'-',A,_‘-.-. o . . A -.-:.-._',.'i’t_;':-:‘-,_s;. AREA 2A - B fa b : OUTER STEEL SHELL DISCONNECT ;.'_:_;,@ / , | FLANGES h _ ®- | EXPANSION JOINT LATCH STOP \\ CAPSULE GUIDE PUMP BOWL CONTAINMENT Fig. 5.7%. VESSEL SHELL Radioactive Material Sampler. 105 Interim waste storage vessels are located adjacent to the processing area in a large canal, To facilitate remote maintenance, vessels are arranged so that all equipment is accessible from above, and all process and service lines can be connected remotely. Over-all building space is dictated by remote maintenance considerations rather than by actual vessel size, 5.2.14 Plant Layout In order to establish uniformity in cost estimation of nuclear power plants, the Atomic Energy Commission has specified certain ground rules (52) covering topography, meteorology, climatology, geology, availability of labor, accounting procedures, fixed charge rates, etc. These ground rules were used in this study. Advantage was taken of a design study and operating experience with a remotely maintained radioactive chemical plant reported by Farrow (32) to obtain over-all plant arrangements, as shown in Fig,., 5.8. The hypothetical site location is 500 miles from the reactor site, The plant is located on a stream that is navigable by boats having up to 6-ft draft. There is convenient highway and railroad access. The plant is located on level terrain in a grass-covered field., The earth overburden is 8 ft deep with bedrock below, 5.2,15 Capital Cost Estimate The cost of the fuel processing plant was apportioned among three principal categories: building costs, process equipment costs, and auxil- iary process equipment and services costs. The building costs included such items as site preparation, structural materials and labor, permanently installed equipment, and material and labor for service facilities. Proc- ess equipment costs were calculated for those tanks, vessels, furnaces, and similar items whose primary function is directly concerned with process operations. Process service facilities are items such as sampling facili- ties, process piping and process instrumentation which are intimately asso- ciated with process operations. 106 . L ol { 560" & Watiimy To Opusati Caaus > { 0] | ¥ i 7 ¢ [ 9.’? g e 2VQ ] T @ & CGyed Srenmace] - \}‘ \ Pu Decay Stomass | o & Coowr. TRaps. @ % ] l > & 0 , A o . Sawvpns Gawamy Evavaram] Lan T Couyno., Koo {Ovam. Couymon Poowm} ¥ AsALyTIEAL = : . - 4 = o \fn ¥ o= CATaE AN Nanidsige . = ) N \/] \Ya LADVES e Oppen Rearm - 1 . : IL TR | v i a I InsTR Suen Oppica Offics » Grount Froort Puaw ‘B -t IR tTet T = bt 'G‘T f | 9 ! . e e . ///// . S B ciwiim. "StoRALL] Crad Covo Trany - Erbasm AT oS Po DAy ~hrpamAak .“, ,, Rebuation Pusc. Mice. i RYAgsniien] . // ; : ._‘_ i 4 ] | a Repn. Elrvaten i Coapitvien . \ CLE MW Awuiy ll 'b‘-’- ‘el 'D‘u-l- t I E l Passvunt Ficom PoAw z i = T'ig. 58‘ Fluoride Volatility Process Plant Layout. *) ', ‘e 107 Accounting Procedure. — The accounting procedure set forth in the Guide to Nuclear Power Cost Evaluation (52) was used as a guide in this estimate, This handbook was written as a guide for cost-estimating reactor plants, and the accounting breakdown is not specific for a chemical proc- essing plant. Where necessaryithe accounting procedures of the handbook were augmented, Process Equipment. = A large number of process vessels and auxiliary equipment in these plants is similar to equipment previously purchased by ORNL for the fluoride volatility pilot plant for which cost records are available, Extensive use was made of these records in computing material, fabrication and over-all equipment costs. In some cases it was necessary to extrapolate the data to.obtain costs for larger vessels, Items that were estimated in this'manner-include the fluorinators, furnaces, NaF absorbers,'and CRP traps. The cost of the UFg-to-UF, reduction unit was based on a unit described by Murray (70). The unit had a larger capacity than was needed for these plants, but it was assumed that the required unit would have about the same over-all cost. Refrigeration equipment and cold traps were estimated from cost data on ORGDP and ORNL equipment, For vessels and tanks of conventional design, the cost was computed from the cost of materialp(INOR—B for most ressels) plus an estimated fab- rication charge, both charges being based on the weight of the vessel., A summary of values used in estimating'process vessels by weight is given in Table 5.1. Some”items of processrequipment were of special design and sig- nificantly different from any .vessels for which cost data were available, For the shells of the prefluorination storage_tanks, the high fabrication cost values shown were obtained by comparison with an available shop esti~ mate for a similar vessel., . ,{ Auxiliary process items such as process piping, process electrical service, 1nstrumentation, sampling connections and their installation were not considered in sufficlent design detail to permit direct estimation. A cost was a531gned to these items which was based upon previous experience - in d351gn and cost estimation of radiochemical proce351ng plants, 1In 'aSSLgnlng these costs cognlzance_was taken of the fact that the plant is remotely maintained. 1108 Table 5,1, Vessel, Pipe, and Tubing Costs ~ INOR-8 ~ Alloy Stainless $/ft $/1b 79-4 Steel 304 Metal Cost | | 3,00 2,66 0,65 Fabrication Cost: | Prefluorination INOR transfer 3,50 - - tanks 3 and 4 ' . Fluorinators ' e 4,00 - Transfer vessels - 3,50 - - HWaste storage vessel ) - = - 2,50 Waste storage thimbles | - - 1.85 UF, dissolvers ' ' 3,50 - - 1/3 in. OD x 0,042 wall tube 6.06 - 26,40 = - ‘1 in. IPS, Sch, 40 pipe 30,05 16.04 - ' - 1-1/2 in, IPS, Sch, 40 pipe #1,67 13,71 - = Buildings. — The building estimate shown in Table 5.2 included the cost of land acquisition, site preparation, concrete, structural steel, painting, heating, ventilation, air conditioning, elevators, cranes, service piping, laboratory and hot cell equipment, etc. The individual costs were calculated using current data for materials and labor, and were based on the drawings prepared. | Process Equipment Capital Cost. = Process equipment capital costs for the two fluoride volatility plants are presented in Table 5.3, These costs are the totals of material, fabrication and installation charges, Building Capital Cost. = As mentioned above, process equipment and buildings were the only items considered in sufficient design detail to permit direct estimation. The remainder of the capital costs were esti- mated by extrapolation from pfevious studies of radiochemical processing plants, The fact that the piant is remotely maintained was an important factor in estimating process instrumentation and electrical and sampling connections, These items are expensive because of counterbalancing, spacing, and accessibility requirements, | Construction overhead fees were taken at 20 per cent of direct materials and labor for all buildings, installed process equipment, piping, ik oy 109 Table 5.2. Fluoride Volatlllty Processing Bulldlng Costs 30 ft3 /day Plant Capacity Materials Labor Total Receiving Area ‘ | Excavating and backfill $ 38,000 $ 18,000 § 56,000 Concrete, forms, etc. 108,000 133, 000 241,000 Structural steel, etc. 58,000 49,000 107,000 Roofing 13,000 ‘16,000 . 29,000 Services 63,000 39,000 102,000 $ 535,000 " 1) ‘Processing Cells $ 85,500 $ 269,700 Excavating and backfill $184,200 Concrete, forms, etc. 520,000 639,000 1,159,000 Structural steel, etc, 277,000 235,000 512,000 Crane area roofing, 64,900 75,600 140,500 painting, etc. . Crane bay doors 390,000 160,000 550,000 Services 301,000 -188, 000 489,000 Bldg. movable equipment 865,000 255,000 1,120,000 Viewing windows | 40,000 2,000 42,000 $4,282,200 Waste Storage : Excavation and backfill $ 95,000 S 44,400 $ 139,400 Concrete, forms, etc. 332,000 496,000 = 828, 000 Structural steel, etc. 400,000 404,000 804,000 Crane area roofing 86,500 100,800 187,300 ~ Painting 4y 500 ¢ouh 500 89,000 Services 565,000 289,000 854,000 Bldg. movable equlpment 225,000 30,000 ~ 255,000 | . $3,1ss,7oo Operations and Laboratories o . - Excavation and backfill § 72,200 $ 33 400 $ 105 600 ~~ Concrete forms, etc. 82,600 115, 100 197,700 - Structural steel, etc. 204,700 43,200 - 247,900 - ‘Roofing 8,500 .4,300 ' 12,800 Super structure 79,100 27,100 106,200 - Misc, structural materzal 31,900 35,100 67,000 Services , 352,000 276,000 - 628,000 - -Misc. equipment - 300,000 43,500 343,500 $1,708,700 110 Table 5.2, Continued Materials Labor - - Total Outside Utilities Cooling tower, motors, $ 70,000 ‘pumps, piping f Water resevoir, pumps, 300,000 piping ' | - ‘Fire protection (house & 35,000 equipment) | Yard lighting 5,000 Boiler house steam heating .300,000 (4,000 kw at $75/kw) Air compressor system 10 000 Steam distribution £ con-. densate return 3,500 Cooling water supply & 40,000 return , Water supply connection 1,700 Process drain lines 3,500 Sanitary sewer connections 3,700 Radiocactive hot draln connections 12,000 Cell ventilation connec- 9,000 tions to stack Off-gas connections to stack 10,000 Storm sewer system 16,000 Electrical substation & llnes 180,000 (3000 kw at $60/kw) , Stack (200 ft) 50,000 Guard house and portals 5,000 Autos, trucks, crane, bull 50,000 dozer ——— $1,104,400 Land and Land Improvements Land (160 acres at $100/acre) 16,000 Leveling & grading ' 50,000 Topsoiling and seeding - 20,000 Fencing, (2 miles at $u/ft) 44 000 Railroad spur, 100 ft 20,000 Asphalt roads & parking areas 200,000 $ 350,000 O %) x) 111 Table 5,3. Installed Cost of Fluoride Volatility Process Equipment 30 ftS/day Plant Capacity Number Description Cost Receiving Cooling jackets for 160 2,75 ID by 15-ft high, $ 120,000 shipping tanks carbon steel Instrumentation Thermocouples, radiation 480,000 monitors, etc. Prefluorination Storage Transfer tanks 2 2-ft x 2-ft; INOR-8; 100,000 0,375 in. shell, 0.5 in. head Furnace 2 2.7 £t x 3 ft; 50 kw 7,000 Fluorination Fluorinators 5 1,75 ft 4. by 9 ft high; 40,000 6 £t3 salt; Alloy 79-4; 0,5 in, shell, 0.5 in. head Furnaces 5 2.7 ft d. x 4 ft hy; 75 kw 16,000 Movable bed absorber 5 '6in.dxu4 fth 25,000 Absorption Absorbers with furnaces 30 6 in. sch., 40 pipe, 6.3 ft 150,000 long Cold traps 15 -40°C units, copper 112,500 15 = =75°C units, copper 37,500 NaF chem, trap 5 6 in, sch, 40 pipe x 6 ft; 3,000 | ~ heated | Vacuum pump 1 50 cfm displacement 3,000 Reduction and Compounding } Reactor - 110 kg/day capacity; Inconel 66,150 Dissolver 2,7 £t d. x 2.7 £t hy 12 ££5 5,500 ~ salt; INOR-8; 0,5 in, shell Heater ‘3.4 ft d, x 3.7 ft h; 71 kw 6,000 Salt make-up tank 2 3.4 ft d. x 6,7 ft h; INOR-8; 26,000 o f£t3 capacity B Heater 2 4,1 ft d. x 7.7 £t h; 178 kw 34,000 112 Table 5.3, Continued Number Description Cost Transfer Tanks 2 4,5 ft d. x 4.5 £t hy 60 ft? $ 100,000 of salt; INOR-8 with heaters Waste Storage Shipping and storage As 2 £t ID by 8 ft h;astainless Included tanks Needed steel 304 L; 10 ft© salt; with op- 0,25 in., shell and head erating expenses Thimbles 1200 2,75 £t 4, x 15 ft h; or- 600,000 . dinary steel 304 L; 0.1875 shell Miscellaneous Equipment ’ Refrigeration unit 2 50,000 Btu/hr at ~u40°C 10,000 Refrigeration unit 2 8,000 Btu/hr at ~-75°C 10,000 HF disrcsal unit 1l 2,8 ftd, x5.3 fth . 500 F2 su-ly system 5 Tank and trailer 35,000 Total Process Equipment $1,987,150 instrumentation, electrical and other direct charges. This rate is higher than current charges for this type of construction and estimates, Archi- tect engineering and inspection fees were taken as 15 per cent of all charges including construction overhead. A summary of the capital cost estimate is presented in Table 5.4, ) 113 Table 5.4%. Summary of Capital Cost Estimate for Molten-Salt Reactor-Fuel Fluoride Volatility Processing Plant Capacity - 30.ft3 of Salt/Day Receiving Area $ 535,000 Processing Cells 4,282,200 Waste Storage 3,156,700 Operations and Laboratories 1,708,700 Outside Utilities 1,104,400 Land and Improvements 350,000 Process Equipment 1,987,150 Process Piping 320,000 Process Instrumentation 205,000 Process Electrical Connections 39,000 Sampling Connections 30,000 e ——— Total, Installed Equipment and Buildings $13,718,150 General Construction Overhead at 20% subtotal 2,743,660 Architectural Engineer, etc,, at 15% subtotal 2,469,294 Contingency at 20% subtotal 3,786,250 Interest During Construction, 9.3% of subtotal 2,112,728 Total - $24,830,082 5.2.,16 Operating and Maintenance Cost Estimates The manpower requireménts were estimated consistently with the proce- dures outlined in the Guide (52); the results are listed in Table 5.5, Materials, utiiities, maintenahce materials, etc. were estimated by consid- eration of the process steps invqived; the estimates are listed in Table 5.6 together with a summary of the labor cost. The total operating ' cost was estimated to be $4,040,850 annually., - 5.2,17 MSCR Irradiated Fuel Shipping Cost The shipping cask must acdommddate a molten-salt shipping cylinder Vhaving a'vplume of 10 ft3; The cost was gstimafed from-themweight which ‘was determined by the shielding reQuirements, A unit cost of $1.00/1b fabricated was allowed, inéluding charges for an INOR liner and a condens- ing-water radiator. - Three casks were allowed so that one might be at the reactor site, a second at the processing plant, and a third in transit. Results are listed in Table 5.7, 114 Table 5.5. Operating Manpower Estimates for ! 30 £t3/day - Fluoride Volatility Plants No Cost ° ($/year) Management | Manager 1 18,000 Assistant manager 1 15,000 Secretary 2 10,000 y 43,000 Production - Superintendent 1 - 12,000 Shift supervisor 4 30,000 Operator 12 66,000 Helper 12 60,000 Secretary 2 9,600 31 177,600 Maintenance Superintendent 1l 10,000 Mechanical engineer 2 16,000 Mechanic 12 69,600 Machinist 3 18,000 Instrument man 8 46,400 Clerk’ 1 4,350 Storeroom keeper 2 8,700 29 173,050 Laboratory Supervisor 1 8,000 -~ Chemist 6 39,000 Technician 10 52,000 Helper 6 28,800 23 127,800 Health Physics Supervisor 1 8,000 Monitor 4 20,800 Clerk 1 4,000 Records keeper 1 3,600 7 36,400 O " * o 115 Table 5.5, Continued | No Cost * ($/year) Accountability Engineer 1 7,000 Clerk 1 4,000 2 11,000 Engineering Mechanical engineer 2 16,000 Chemical engineer L 36,000 Draftsman 3 15,900 Secretary 1 4,500 10 72 ,400 General Office Manager 1 5,000 Accountant 1 4,800 Payroll clerk 2 8,000 Purchasing agent 1 4,800 Secretary 2 8,000 7 30,600 Miscellaneous o Guard 8 32,000 Fireman 4 16,000 Receptionist 1 4,000 Laundry worker 3 10,800 Nurse = - 1 4,800 Janitor 3 10,800 - 20 . 78,400 Total e 133 750,250 116 Table 5.6, Fluoride Volatility Plant Direct Annual Operatlng Cost Shipping - Storage Tanks (50 at $2,500) $ -125,000 Chemical Consumption Fluorine (at $2.00/1b) 120,000 KOH (at $0.10/1b) 16,600 Hydrogen (at $2.00/1b) 4,500 NaF (at $0.15/1b) 3 1300 Nitrogen (at $0.05/f£t%) 3,400 HF (at $0.20/1b) 6,100 Graphite (at $0.15/1b) 1,100 Miscellaneous 5,300 157,300 Utilities Electricity (at $0.01/kw hr) 362,000 Water (at $0.015/1000 gal) 5,700 Heating (based on steam at $0,25/1000 1lbs) 8,500 376,200 Labor® Operating 406,400 Laboratory 127,800 Maintenance 173,050 Supervision 43,000 Overhead (at 20% of above) 150,000 900,250 Maintenance Materials Site b 10,000 Cell structures and buildings 76,000 Service and utilities 78,800 Process equipment® 243,300 482,100 Total Direct Operating Cost $2,040,850 qSummarized from Table 5.5. bBuilding services excluded. c 2 > * Includes process equipment, process instrumentation and sampling. L Table 5.7, MSCR Irradiated Fuel Shipping Cask Data and Shipping Cost™® Cask weight 100,000 1bs Cost of cask $100,000 Number of casks 3 Salt volume in shipment 10 £t3 Age of salt at shipment 20 days Days salt accumulation in shipment 6 days Round trip distance 1000 miles Round trip time 10 days Method of shipment Rail Number of shipments per full power year 50 Freight rate $2.40/100 lbs- 1000 miles Unit shipping cost $330/£t3 salt *Data and cost adapted from Reference 20, 5.2.18 MSCR Unit Processing Cost The various bases and contributions to the unit processing cost are collected in Table 5,8 for the fluoride volatility central plant process- ing of MSCR fuel. For the feference design fuel containing 32 kg of thorium per fta, the unit cost was $36.90 per kg of thorium, Although the cost was expressed in terms of $/kg of Th, it should be remembered that only isotopes of uranium are recovered. Stripped salt, containing valuable thorium, lithium-7, and beryllium, as well as fission prdducts; is discardéd;"Additional.br alternate processing would be re- . quired to recover any of these components., 118 Table 5.8. Unit Processing Costs, Central Fluoride Volatility Processing Plant for MSCR Fuel Capacify of plant | 30 ftalday Reference design fuel 32 kg Th/fts Annual charges: | | Capital (24,8 million at 15%) $3.7 million Operation and maintenance $2.0 million Daily charge (80% plant factor) $19,500/déy Batch size : 188 £t3 or | 6000 kg Th Processing time 6.3 days Turn-around time 7 2 days- Processing plant cost $26.60/kg Th Shipping cost - $10.30/kg Th Total processing cost $36.90/kg Th 5.3 Thorex Central Plant The reference plant described in the Guide (52) is a central facility capable of processing 1000 kg Th/day with thorium discard or 600 kg Th/day with thorium recovery. The plant was designed specifically for thorium metal or thorium oxide fuels; however, since other types of thorium fuels were not specifically excluded, it was assumed that the piant would also accept a fluoride-salt fuel, It was further assumed that the fluoride fuel would be processed at the same base charge as the metal or oxide fuel. This assumption amounted to assigning the same charge to a fluoride head- end treatment as to dissolution and feed preparation steps for the other fuels, The tail-end treatment for the conversion of Thorex nitrate product to fluoride feed material was assigned a cost that was thought to be repre- sentative of the processing steps. o D bp! 119 5.3.1 Head-End Treatment The head-end treatment shown schematically in Fig. 5.9 has not been demonstrated, However, the chemical principles have been established (40) by laboratory investigations of the stability of fluoride salt fuels., It is known (40) that the oxides of uranium, thorium and beryllium are very stable compounds having the indicated order of stability U0, > BeO > ThO, and that oxygen or oxygen-bearing compounds must be eliminated from fluo- ride salt fuels to insure their stability, In the proposed head-end treat- ment, the draw-off from the reactor at 500 - 550°C would be contacted in a spray with steam or high temperature water (v200°C) to precipitate the oxides of uranium, beryllium, thorium, protactinium and some of the fission products., It is believed that rare earth oxides can be precipitated in this manner. Lithium fluoride is a very stable compound and would probably not enter into reaction with water. It should remain in the system as LiF and be frozen into small crystalline particles. The hydrolysis would fofm large quantities of HF which in aqueous media is rather corr?osiire° Therefore the selection of materials of con- struction for the precipator will be a pr-bblem° Dispbsal of HF can be accomplished by dilution with large volumes of air and dispersion from a stack or by neutralization with an inexpensive base. Some cleanup of the HF stream will be required because of volatile fission products. Reuse of this HF in the subsequent hydrofluorination step (see Fig. 5.8) may not be ~ feasible because of water vapor in the gas. The second step in fhé’head-end treatment is dissolution of all the hydrolyzed components that are soluble in nitric acid, The oxides of uranium, thom.ums protact;nlum and rare earth products should dissolve quite readlly, Since lithium fluor;de and beryllium oxlde are quite in- | soluble in aqueous medxa,_neglxglble amounts of these compounds should be - dissolved. Also, it is almost certain that some of the fission products will be insoluble and remain with the lithium and beryllium. Dissolution 'should proceed smoothly because it has been shown by Pitt (73) that the particle sizes produced when a molten salt is sprayed into water are in the micron and submicron range, 120 off-gas HNO, HF 1 - Molten Precipitation Dissolution | Filtration Aqueous FP's ~200°C 100°C UOx(NOy), - Th (NO,), {FP).(NO. #,0 or Steam : l ) LMoy, 200°C | Soiid Product LiF BeO Make - Up { FPS) LiF-Be F'-Th F.-UF. Waste Discard HF, H,O,H, ‘ Denitration Reconstituted | Hydrofluorination LiF Be Fg This UF, HF+H, Fig. 5.9. Nitrogen Oxides Th{NO,}- ThO, ~ 400°C Solvent Extraction Waste FPs — ORNL-LR-DWG 71093 Th(NO, ), Product Steam Proposed Aqueous Processing for MSCR. U0,(NO;), Product £ 121 Dissolution is followed by solid-liquid separation either by filtra- tion or centrifugation. Aqueous nitrate solutions are fed into a feed adjustment step preceding solvent extraction by Thorex and the insoluble material is routed to hydrofluorination., At this point a portion of the solids can be discarded as a purge of fission products that remain in the precipitate. 5.3.2 Solvent Extraction Decontamination of thorium and uranium can be accomplished by well- established Thorex procedures. Aqueous nitrate solutions are evaporated until about 0,15 N acid deficient and fed to an extraction column. In the extraction column both thorium and uranium are extracted into an organic phase (tributyl phosphate) leaving the bulk of the fission products in the " aqueous phase, with decontamination factors up to 105. Waste from the first extraction cycle contains all of the protactinium that was in the feed stream. However, the amount is insufficient to war- ‘rant recycling the waste after an additional decay period. In the interim between discharge from the reactor and chemical processing, the fuel ages about 120 days so that only about 15% of the protactinium remains undecayed. The waste is given permanent'storage in large underground tanks., In a fuel recycle system such as the MSCR it is not necessary to de- ~contaminate further by the use of additional extraction cycles. The pres- ence of 232U and 228Th will make recycle fuel too radiocactive for direct handling, regardless. After extraction, therefore, it is sufficient to paftition-uraniufi andtthorium in a stripping'column_by the proper adjust- ‘ment of organic and aqueous flow rates. In this operation, thorium is stripped from the_organié phasé into an aqueous phase; uranium remains in the orgénic“phase,r A-subsequept-stripping operation returns the uranium to the aqueous phase. _The'prdduce:streams are'respectively Th(NO,), and U02(N0 ' 3)2;-” ' 5.3;3 Tail-End Treatment Fuel reconstitution begins with acceptance of the nitrate products from the solvent extraction plant, It is necessary to convert the nitrates it 122 to the fluorides. In the case of thorium this is accomplished by pre- paring the oxide in a denitration process followed by hydrofluorination in a molten salt mixture, The steps are as follows: steam denitration HF 1-—--—-——” Th(NOa) “400°C = Th02 in molten salt™ ThF (nv600°C) Steam denitration is an established procedure in the sol-gel process (33) for preparlng highly fired, dense ThO2 or ThO2 UO2 fuel, In this case the aqueous nitrate solution from the Thorex process would be evapo- rated to crystallize Th(NOS)u; superheated steam for the actual denitration. Final preparation of the the crystals in turn would be contacted with fluoride has to be accomplished in a second high temperature operation in -molten fluoride salts. Thorium oxide is quite intractable to attack by hydrogen fluoride under most conditions; however, in the presence of molten fluorides the reaction will occur., The presence of other high valence com- pounds, e.g., other thorium or uranium fluorides, in the melt abets the dissolution. The conversion of uranyl nitrate to the tetrafluoride is not as straightforward as that of thorium because of the required valence change, In the Excer process developed at ORNL, uranyl nitrate from the last Thorex stripping column can be fed directly onto a cation exchange resin (Dowex 50 W) which absorbs the uranyl ion. After loading, the resin is eluted with aqueous hydrofluoric acid to produce UOze, which is reduced in an electrolytic cell. The aqueous solution is allowed to flow into a mercury cathode in which UF, ¢ 0.75 H,0 precipitates from the aqueous phase. The precipitate is separated by centrifugation or filtration and dried. Water of hydration is not tenaciously held, and moderate drying conditions are sufficient to expel it. A second method of converting 002(N03) to UF, is to reduce U(VI) to U(1IV) in the presence of fluoride at 500 - 600°C, In this tail-end treat- ment the two nitrate solutions of thorium and uranium are mixed and co- denitrated using superheated steam to yield ThO2 and 003° The mixed oxides are converted to the fluorides in a 500 - 600°C molten fluoride salt bath (} 123 by contacting with a gaseous mixture of H, + HF. Uranium is reduced by the hydrogen and hydrogen fluoride to theztetravalent form and dissolves as UFu; thorium is metathesized to Tth which also dissolves in the melt. The processing cycle is completed when the uranium fluoride is mixed with fluorides of thorium9 lithium, and beryllium oxide plus make-up feed and hydrofluorinated. Beryllium recycled as Be0O dissolves as BeF, when 2 hydrofluorinated in molten fluoride solutions, 5.3.4 Processing Costs The unit processing costs were computed according to the prescription given in Guide to Nuclear Power Cost Evaluation, Vol., 4, Section 460, The escalated daily charge for operation of the Thorex plant was taken as $17,500 in late 1962, Since the thorium was recycled,'the capacity was 600 kg/day. As described in a previous section, the daily increments of fuel withdrawn from the reactor were accumulated for 120 days and combined into a single processing batch, If the withdrawal rate exceeded 1.25 ft3/day (40 kg Th/day), the batch size exceeded 4800 kg of Th and re- quired eight days or longer to process, For this range of process times, the "turn-around-time" is specified to be eipght days in the Guide, Thus . the total processing time is 120r/600 + 8 days (where r is the withdrawal rate in kg Th/day) and the total amount processed is 120r kg of Th. The cost for operating the Thorex plant is thus - 17,500 [wc-“o” + 8]' Separation Cost o - ' 120r (29.2 + 1167/r)-$/kg Th '1’_Ihe'c05f df féducing séiifi fuel elements into é:form suitable for pro- cessing (aQuebus soifition Of'ufianylfand thorium nitrates) is included in the daily'operéting cost (io#);"itseEmsAlikely that the cost of convert- ihg the MSCR fuel by'the method pfiopoéed in Sec, 5.2 would be less costly than the reduction of solid fuel elements by Darex or Zircex. However, an investment of $500,000 for MSCR fuel head-end treatment was allowed, based 124 on estimates extracted from reference 20, This resulted in a head-end treatment charge of $0.53/kg Th processed. The cost of converting recovered thorium nitrate to ThFu was not given in the Guide. However, the cost of converting low enrichment uranium to UF6 was only $5.60/kg. Since the proposed thorium conversion process is simpler and shorter than the uranium process, it seemed adequate and con- servative to assign a cost of $5.00/kg Th for the conversion of ThE,,. Summing these charges, one has Processing Plant Cost;= (34,7 + 1167/r) $/kg where r = kg Th/day removed from the reactor. Shipping costs are presented in Section 465 of the Guide (52), and include freight, handling, cask rental, and property insurance. The cost is $16/kg of uranium (or thorium) for spent fuel elements, MSCR fuel must be packaged before shipment. 1In the scheme proposed, the daily productions of 1.67 ftalday are accumulated in 35 ft3 batches in the spent fuel facility (Sec. 4.8.10) which contains 5 tanks each having a capacity of 35 ft3 and provided with sufficient cooling to remove afterheat from freshly irradiated fuel., A 10 ft3 batch is drained into an INOR "bottle" (cylinder) one foot in diameter and 13 feet long. This is in- serted into an individual shipping cask and transported to the Thorex plant, If the bottles cost as much as $10,000 apiece, but are re-usable for at least 10 years, they will add to the shipping cost only $1,74/kg Th at the lowest rate of processing (40 kg/day). The charge for shipping decontaminated thorium back to the plant is represented to be $1.00/kg in the Guide. Since the processing plant operating charge was all levied against the thorium, the only charges on the uranium are shipping at $17/kg per round trip and reconversion, for which operation the Handbook gives $32/kg for converting the nitrate of highly enriched 235y to UFG, Although 'the‘cost of conversion to UFu should be less, the given cost was assumed. The total charge for the uranium was thus $49/kg, and was charged against all isotopes of uranium, including 238y and the precursor, 233pa, ” 125 S5.u Comparison of Processing Cost Estimates For the purposes of comparison, the results of three other processing cost studies are cited. The first, by Carter (20), dealt with a central Fluoride Volatility plant designed especially for processing MSCR fuel at a rate of 31.5 fta/day (1000 kg Th/day). The estimates were based on the same design study (21) used to prepare the estimates reported in Sec., 5.2, The scaling, however, was performed on the plant as a whole, rather than with individual items, and the scaling factor was 0.31. The estimated cap- ital cost was $31.5 million, and the annual operating expense was $3.0 million. This study is referred to below as the "CMS" study. The second study, also by Carter (20), dealt with a central Thorex facility adapted to process fluoride fuels (1000 kg Th/day) and was based on a detailed analysis of the head-end and tail-end processes (described in the previous section) as well as the Thorex separation process, and was based on an unpublished design and cost study performed by Carter, Harrington, and Stockdale at ORNL in which flow sheets, equipment specifi- cations, plant arrangement draw1ngs building drawings, etc, were prepared and the items were costed. For brevity, this study is referred to below as the "CHS" studyo It'iS, essentially, an lndependent estimate of the facility assumed in the Guide to Nuclear Power Cost Evaluation (52), “adapted for fluoride fuels. The estimated capital oost, scaled to a capac- ,lty of 1000 kg Th/day, was $36 5 mllllon, and the annual operating expense was $3.6 mllllon._tf. : | | - o The thxrd study, performed by W, H Farrow, Jr, (32), dealt with sev- eral radxochemlcal separatlon plants for several dlfferent solid fuels and rfclads and with both dlrect and remote malntenance, The purex process, which is simllar to Thorex, was employed for the separations, The most applicable case was that of a remotely maintalned plant capable of treatlng one tonne '"per day of natural uranlumoi Although the process was descrlbed in consid- ':erable deta11 a cost breakdown was not given by Farrow, The capital cost ‘was $43 mllllon, and the annual dlrect operating cost was$3‘7inillion.' One mlght reasonably assume that a Thorex plant of the same capac1ty would have approx1mately the same costs, and that the head and tail-end treatments for fluoride fuels would be no more expen31ve, ‘and perhaps less, than.those of the solid fuels, 126 It is seen that the estimates of the total cost, including shipping, - vary by a factor of two, ranging from $37‘tq $75/kg of thorium processed. It seems plausible that processing MSCR fuel' in a-central Fluoride Volatil- ity facility will cost less than $50/kg, and possibly less than $40, and that processing in a Thorex plant will cost less than $70, and possibly less than $50 per kg of thorium, | : | | Table 5.9, Comparlson of Estimates of Proce351ng Costs for the MSCR Reference Design?® This Work CMS ~ AEC CHS Farrow Type of Plant Fl. Vol. Fl, Vol. Thorex Thorex (Purex) Location Central Central Central Central Central Capacity, kg Th/day 1000 1000 600 1000 (1000)P Batch Size, kg Th 6000 6000 6000 6000 6000 Turn-Around-Time, Days 2 -2 _ 8 2 - (2)¢ Capital Cost, $10° 25,2 31.2 - 36,5 43 Annual Operating & main- 2.0 3.0 - 3.6 . 3,7 tenance Cost, $10° | - Process Plant Cost,d 26,6 3,5 56,5 40.5 46.4 $/kg Th ‘ , . s f f f c Shipping Cost, $/kg Th 10,3 10.3 18.7 10,3 (20) Processing Cost, $/kg Th 36.9 44,8 75,25 50.8 (66.4)° aSpent fuel is withdrawn from the reactor (1.67 ftalday; 53.3 kg Th/day) and shipped to a central processing plant in 10 £t3 batches. These are accumulated in 6000 kg Th batches, cooled for an average of 90 days, and processed in six days (10 days in AEC plant). , bPurex plant capacity of one tonne of natural uranium per day ~ assumed equivalent to one tonne thorium per day. ®These items were not estimated by Farrow; values chosen were thought to be consistent with his general treatment. dFixed charges were 14.46%, except for AEC plant where'daily oper- ating charge (escalated to 1962) was given in reference 52. ®Turn-around-time was'B'days. With a more realistic time of 2 days, and shipping costs of $10.30/kg Th, the total costs would be only about - $50, which is comparable to the other estimates listed, except Farrow's. frable 5.7. ' ) w .A. e i 1] o A 127 6. FUEL CYCLE ANALYSIS 6.1 Analysis of Nuclear System The nucléar calculations for the MSCR were performed by means of the MERC-1 program for the IBM-7090, This program, described by Kerlin et al., in Appendix D, uses the mnltigroup neutron diffusion code Modric and the isotope code ERC-10 as chain links, 6.1.1 Computer Pbograms Modric., ~ This program was employed in a 34-group version using the group energies and cross sections of the various elements given in refer- ence 54, The cross sections were adopted for the most part from Nestor's tabulation (72), with some minor modifications described below, Maxwell- Boltzman averaged thermal cross sections and resonance integrals of impor- tant materials are listed in Table 6.1. Thermal spectrum hardening was ignored, Although the treatment of downscattering in Modric provides for the transfer of neutrons from any group into any of the ten next lower groups, the required scattering matrices were not available when the MSCR calcula- tions were begun. Subsequently, the matrices were computed taking into account fast fissions and inelastic scattering in thorium and the fact that the elastic scattering lethargy decrements were in some ranges larger than the group widths., A single caloulation was made to defermine the impor- tance of treatlng the downscatterzng in thls more precise manner. The effect was found to be 1ns;gn1f1canto _ ERC-lOO_e-The baSLc 1sotope equatlon in ERC-10 computes the concentra- tion that is in equlllbrlum Wlth the sources (make-up, recycle, transmuta- tion, decay, f15510n) and the sinks (transmutatlon, decay, waste, sales, 'f_recycle)o. One isotope (usually 233y op 235J) may be selected to satisfy a crltlcallty equatlon- the feed rate or sales rate required to maintain the eritical concentratlon is then computed Three isotopes, 23“0 236U “and 2380 approach equllibrlum with periods long compared to the assumed fuel life (30 years). The equilibrium 128 Table 6.1, Cross Sections and Resonance Integrals Used in MSCR Multigroup Neutron Calculations (Values in barns per atom) Mateprial ) Thermal Cross Section? | i Resonapce_;ntegral§b | Fission (vof) Absorption (aa) Fission (vof) Absorption (oa) 2321y 8,77 8.286 x 1071 9,684 x 10 233pa ' 1.8943 x 10 8,67221 x 10° 233y 6.54 x 102 3,40 x 10 12.01972 x 10° 9.60119 x 10° 234y 5,54 x 10 - 1,0602 x 10 6,8%u453 x 102 235y 6.29 x 102 3,05 x 102 7,54945 x 10> 4,66792 x 10° 236y 2,10 5,487 2.87528 x 102 237Np | : 1,07 x 10 1.0983 x 10 5,70047 x 10° 238y | 1.3 3,225 2,77241 x 10° 6Li 4,720 x 10° 4,58811 x 10° 7 1.66 x 1072 | 1.5774 x 1072 9Be | 5,048 x 10~3 4,748 x 1070 Log 4,5 x 1077 | 1,975 x 1070 12¢ | 5,048 x 107° 1.897 x 107% INOR-8 2.874 . ) 6.004 1353 1.60 x 10° 4.5756 % 10° ®Maxwell-Boltzman averaged at 1200°F, bCut-off at 0,437 ev.; infinite dilution values. concentrations of these were not computed; instead, first approximations of their time-mean concentrations (starting_with'a reactor initially inven- toried with 235y 95% enriched) were computed as described in Appendix H. The transient behavior of all other isotopes, including 233pa, 233U, and fission products, was ignored in the optimization studies, The afiprox— mations involved in this approach are examined in Appendix H, ) MERC-1, = Input (5u4) consists of specifications of the_geomefry and dimensions of the reactor regions, initial guesSes‘at'the composition, in- formation on power, fuel volfime,”processing_modes and rates, éomposition of make-up materials, unit values, and processing costs. The output consists ) 129 of nuclear data (breeding ratio, mean eta, neutron balance, etc.), process- ing data (feed and produotion_rates),rand a fuel.cycle cost (inventory, make-up processing, etc.). An examination of these data discloses the principal items of cost and suggests changes in specifications which might reduce these. In some cases, the effect of changing an input parameter is not readily predietable; e.g., increasing the concentration of thorium in the fuel stream usually increases the conversion ratio and so reduces the 23%) feed requirements but on the other hand increases the fissile inven- tory. In cases such as this,'the input parameters are varied systemati- cally in a "factored" set of calculations yielding the maximum information from a minimum number of cases. The fuel cycle cost is thus optimized with respect to several variables simultaneously. (See Section 6.4.1.) 6.1.2 Reactor'PhySics Model The reactor was computed as a homogeneous mixture in equivelant spher- ical geometry (i.e., the input diameter was 1,09 times the diameter of the cylindrical core); Thus'the heterogeneity of the core, which is appreci- able, was ignored. Butathis_treatment is conservative in that the reso- nance escape is underestimated, resulting in a pessimistic estimate of the mean eta of the system. In regard to the estimated captures in thorium, these can always be matched at some neighboring concentration. Aside from a minor effect on the spectrum of neutrons, the chief error introduced is a slight underestimate of the thorium invenfOry, But this is not important, for the 1nventory charges contrlbute only a small portlon of the total fuel costs (less than 5%). ' J ~ The equ;valent spherlcal core comprlsed three zones; an inner zone - con51st1ng of a homogeneous mixture of fuel salt and graphlte, a spherical annulus about one inch thick filled with fuel salt, and a spherlcal reactor vessel one inch th:.ck° The mean, effective temperature of the fuel was 'assumed to be 1200°F, and this temperature was also assxgned to the graph- 1te, which however, may run two or three hundred degrees warmer. '651;3"jcrossTSection.Dataf- Thorium=-232, -~ Saturation of resonances (self-shielding) was found to be important in five of the neutron energy groups. The effective cross 130 o sections of thorium in these groups were calculated by means of a éorrela- gii tion developed by J. W. Miller (Appendix B) and based on the theoretiéal anélysis of effective resofiance ihtegrals made by L, Dresner (29), Protactinium-233, = A 2200 m/s cross section of 39 barns, as recom- mended by Eastwood and Werner (30), was assigned. A resonance integral of 900 barns was adopted. This was distributed as shown in Appendix A, Uranium-233. = A value of 2,29 was adopféd for eta at thermal energies, “based on the recent measurements of Gwin and Magnuson (43), For energies above thermal, Nestor's estimates (72) were used, as tabulated in Appen~. ~dix A. Resonance saturation effects were ignored. | o Beryllium, — The (n,2n) reaction of energetic neutrons in °Be was ig- - nored. It is of small importance in this graphite moderated reactor, and is moreover offset by the disadvantageous (n,a) reaction which uses up a - - neutron and leads to the formation of 6Li, | | Fission Products., - These, excepting xenon and samarium, were handled collectively in the Modric calculation and individually in_the.ERC calcu- lation, as described below. An "effective" concentration of an "aggregate" fission product was computed from ERC results and used in the multigroup calculation in conjunction with an arbitrary set of group cross sections composed of a hypothetical standard absorber having a thermal cross section of one barn and an epithermal cross section éorreSponding to a 1/v varia- tion above thermal. ' ' Thermal cross sections and resonance integrals for fission products were mostly taken from the compilatiofi_of Garrison and Roos (37), supple- - f mented by estimates from Bloemeke (9), Nephew (71), and Pattenden (90), The data used, including fission yields and decay constants, are tabulated in Appendix E. 6,2 ‘Analysis of Thermal andruechanical System The analysis included consideration of fluid flow in and mechanical arrangement of the reactor and equipment associated with the extraction of heat from the fuel stream, as they affect fuel cyc_:lercosts° - L - 131 6.2,1 Maximum Fuel Temperature It is believed that the rate of corrosion of INOR will be very'small and that the heat exchangers'will have a very long lifetime if the temper- ature of the fuel solutioh does not exceed 1300°F (12). The maximum allowable temperature may be higher; if so, future generations of molten- salt reactors will be able to achieve higher specific powers and higher thermal efficiencies, 6.2.2 Minimum Fuel Temperature For this, 1100°F was selected, Earlier work (3) had shown this to be a reasonable value, and calculations summarized in Table 6.2 confirmed its optimal quallty for the MSCR, Perturbations of the affected capital costs were calculated in Table 6. 2, and it was concluded that the optimum temper- ature is very near 1100°F (see also Sec, 3.1.12). 6.2.3 Velocitz Erosion does not appear to be a problem in salt-INOR systems, nor does there appear to be any depeadence of the corrosion rate on velocity. Rather, the velocity appears to be limited by considerations of pumping power and stresses inducedoby'pressure gradients. Pressure drop across the heat exchangers is limited by the rapid increase in cost of INOR shells with increa51ng wall thlckness° Likewise, maximum allowable pressure drop across the reactor core is determined by limitations on strength and thick- - ness of the reactor vessel and internal support members . Velocities, pres- sures, and wall th1cknesses at varlous points in the fuel 01rcu1t of the | reference de81gn are llsted in Table 6, 3. “The pumplng power requlred per heat exchanger loop was. calculated to ’ berabout 1500 horsepower, with a margln of 500 horsepower for pump and f'motor inefflciencles and unforeseen losses, This pumping power requirement ';rresults in a pumping cost (with electric power at 4 mllls/kwhr) of 0,004 mllls/kwhre. | | " | Table 6,2, MSCR Minimum Fuel Temperature Optimization® Minimum fuel temp., °F Fuel stream temp. rise, °F Fuel salt flow rate, lb/hr x 10~ Heat exchanger log-mean temp. difference, °F 6 Heat exchanger area, £t2 Heat exchanger pressure drop, psi Fuel side Coolant side Heat exchahger volume, £t° Fuel side Coolant side Fuel circuit piping Pressure drop, gsi' Pipe volume, ft¥ Pipe weight, lbs. Pumps Fuel circuit, hp. Coolant circuit, hp Net capital cost increment $million Case 1 1050 250 89 144 614,000 (+5477%) 62 101 696 (+$3,025) 584 (+$125) 133 255 (-$1,625) 15,400 (-$31) 9;100 (-$736) 17,300 (+5256) $+1.49 - Case 2 1100 200 111 174 53,000 80 8y 575 488 150 320 18,500 12,800 16,000 Reference | Condition Case 3 1150 150 148 200 46,000 (-$306) 123 73 504 (-$1,775) 424 (-$83) 195 410 (+$2,250) 21,300 (+$28) 22,200 (+$1,872) 15,200 (-$160) $41.83 #Capital cost increments with respect to reference condition (1100°F) associated with each In this calculation, reactor outlet temperature, configuration, heat exchanger cross section, fuel velocity in piping were held constant, as was item are given in parenthesis in $thousands. flow rate of coolant salt and its temperature extremes. and the thermal efficiency was not affected. The power of the reactor was constant, cel i vt st 4) 133 Table 6,3, Characteristics of Fuel Circuit of 1000-Mwe Molten Salt Converter Reference Design Reactor | Minimum Location ‘ - Velocity Pressure Wall Thickness (ft/sec) (psia) (inches) Pump discharge | 35 1190 0,406 Top of heat exchanger - 185 - Heat exchanger tubing | g - | 0.035 Bottom of heat'exchenger 35 95 0.250 Bottom of reactor : - 80 - Fuel channels in core 4,3 - 7.4 - - Top of reactor (Pump suction) 20 22,5 0,312 *Shell-side 6.,2.4 Fuel Volume Contrxbutlons to the volume of the fuel system are listed in Table 4, l for the reference design. The volume of fuel in the external system de- pends on the oower level and various limitations such as those on salt temperature, pbessure, velocity, thermal stress in and minimum thickness of heat exchanger tubing, etc, The reference system was designed with consid- erable conservatism. The fuelrvolume could be reduced appreciably with an increese in specific power; pumpxng power costs and capital 1nvestment in pumps would increéSe., ‘Cost of heat exchangers might also increase. The deszgn of the system should be optlmlzed Wlth respect to the sum of all the - costs affected but this lay outsxde the scope of the present study. The power dens;ty in the external portion of the fuel system of the wjreference des1gn is approxlmately 2.8 th/ft . This is very much smaller ithan the 7.6 th/ft used in a prlor study of a molten salt breeder (3) which was based on a study by Splewak and Parsly (99), who est1mated a sgec1f1c power of 4,9 th/ft for a first generation plent and 7.6 th/ft for subsequent plants, 134 Since only about one fourth of theltotal is contributed by the fuel in the active core, the total fuel volume is not very sensitive to changes in fuel volume.fraction in the core. The total volume is rather"more'sen- sitive to changes in core ‘diameter. The volume of nettron-active fuel in- creases as the cube of the core dlameter (helght equal to dlameter) while volume in the radial annulus and in top and bottom plenums increases as the square. About 40% of the total volume is affected, In a fully optimized system, the fuel volume might plausibly be | 2000 fta, and perhaps be as low as 1800 ft3, provxded some of the holdup in end-plenums, etc. can be eliminated. 6.3 Analysis of Chemical System The composition of the fuel stream as a function of its chemical in- teraction with the reactor environment and with the processing plant is considered. ~Behavior of xenon-135 is important and is discussed. in detail in Section 6.8. Stagnation of fuel in crevices between moderator elements may be important; however, it does not seem.possible at this time to eval- uate the effect except to say that parasitic captures of neutrons in fission products immobilized in such places will take place° This uncer- tainty was lumped with thaf'aesooiated with eorrosion'products as discussed below. | - So far as the composition of the fuel stream is concerned, the chem- ical effects of the two proposed processxng methods (Thorex vs fluoride volat;llty) are the same. The recycled uranium is radioactive and must be handled at least semi-remotely whether the thorium is recycled or not° Both methods result in recycle feed containlng only negligible amounts of nuclear poisons (other than isotopes of uranium) and both return 233pa removed from the reactor as 233U with 1osses that depend only on the hold- ing time prior to chemical treatment. 6.3.1 Thorium-232 ~ Thorium may be recycled if Thorex processing is used; however, accord- ing to the Guide, the capacity of a multi-purpose processing plant would be w ) [ 135 reduced by 40% (52)., (One could design the plant to handle the thorium without loss of capacity, however, and recover the thorium at no extra cost.) It turns out that this reduction in capacity almost exactly offsets the value of the thorium saved. With the fluoride volatility processing, thorium is not recovered except by means of an additional step presently not available, 6.3.2 Protactinium-233 With mean residence times of the order of 1000 days, protactinium for ‘the most part decays while still in the circulating fuel system. However, the process stream carries 60-80 grams per day out of the reactor system. In the proposed reference design processing scheme (Section 5.2), the pro- cessing stream is accumulated for 120 days and then held as a batch for an ~ additional 30 days, giving a total hold-up time of 150 days and an average decay time of 90 days. At the time of processing, 85% of the 233pa in the mixed batch will have decayed to 233U, so that the maximum loss of 233pa will be only 9 to 12 grams per day. 6.3.3 Uranium-233 The loss per pass through the processing plant was assumed to be 0.3%. In the reference designrreactof,'the'product_stream from the process plant was recycled to the reactor; however, the economics of sale of the product stream was also,examinefi.rr 6.3.4 Uranlum-234 " Natural uranium is ‘99, 27 welght per cent 238U 0,72 per cent 235U and 70'0055 per cent 234y, - If there were no enrichment ln 234y relatlve to 2350 a diffusion plant product contalnlng 95 wezght per cent 235y would _ also contain 0, 726 per cent 234y with the balance being 238y, 1In order to - allow for some enrichment, the composxtlon of the feed was taken to be 95 per cent 235y 1.per cent 23“0 and 4 per cent 238y, s with 233y, the 7 process;ng ‘losses were 0.3 per cent/pass° Initially the reactor contains little 23%U, The apfiroach to equilib- rium is slow, and, as seen in Appendix H, the average concentration over . 136 a period of thirty years is only 65 per cent of the equilibrium concentra- tion, The average value, rather than the equilibrium value, was therefore used in the nuclear calculations, 6.3.5 Uranium=235 An enrichment of 95% was selected for the make-up. Losses in process- ing were assumed to be 0.3% per pass. 6.3.6 Uranium=236 “ The concentration of this isotope also approaches equilibrium slowly with respect to a fuel lifetime of 30 years. A concentration averaged over the 30-year life was used. Losses in processing were assumed to be 0.3%, 6,3.7 Neptunium-237 This was assumed to be removed completely in the processing plant. 6.3.8 Uranium-=238 For reasons given above, the fuel make-up stream was assumed to con- tain 4% 238y by weight., Losses in the process plant were assumed to be 0.3%., The 30-year average concentration was used in the nuclear calcula- tions instead of the equilibrium concentration. 6.3.9 Neptuniufi¥239'and Plutonium Isotopes Only small amounts of these will be formed, and they are lost in the. waste. Accordingly, their formation in the fuel stream was ignored and no breeding credit was taken for absorptions in 238y, 6.3.10 ‘Salt - The fuel carrier in the reference design consisted of 68 mole per cent LiF, 23 per cent BeF,, and 9 per gent_ThPu. Lithium in the make-up salt was 99,995% ‘Li. Captures in ’Li, Be, and F were lumped under an equiva- lent isotope "Carrier-1." The mean reactor concentration of 6Li and neutron captures therein were computed separately. Lithium and beryllium in the process stream are lost to the waste, and no value was assigned to *) 4 137 the waste. The make-up rate was made equal to the discard rate for ’Li and Be; the ®Li feed rate was proportioned to the 7Li make-up rate. 6,3.11 Xenon-135 and Related Isotopes In the reference design reactor, it was assumed the graphite has a -5 diffusion coefficient no greater than 107° cm?/sec (D = 10 cm2/sec for - MSRE graphite), a porosity no greater than 0.01, and that 10 per cent of the fuel stream is recirculated to the dome of the expansion chamber and the pump bowls, Here Xe is desorbed and swept away in a stream of helium ~gas with anréfficiency of 100 per cent per pass. With these assumptions, the loss in breeding ratio due to absorptions in 135Ke is 0,017 as shown in Appendix G, where the losses corresponding to other assumptions are also given, 6.3.12 Noble Metal Fission Products It was assumed that this group of isotopes, comprising Mo, Rh, Ru, Pd, and In, "plate out" on INOR surfaces in the fuel circuit with an efficiency of 1.0 per cent per pass. £.3.13 Other Fission Products These are removed 100 per cent in the Fluoride Volatility process, and - only negligible amounts will be present in the recycle stream from a Thorex Plant ° 6.3.14 COPrOSlon Products Data are meager from whlch the concentratlon of corrosion products in the .circulating fuel stream could be estimated. In an 1n-pile loop oper- ated for 15,000 hours, the concentratlons,of_iron,.nickel, afid chromium éppearedtogflupfuatg_aboutequilibrium values (84, p. 79), On the basis of these data one might expect the fuel to contain 50 ppm of nickel, 500 ppm of chromifih,-and'about'250 ppm of iron, In the reference design ,reactor,, these concentratlons would result in a poison fractzon (loss in breedlng ratio) of 0,006 units, A poison fraction of 0,008 units was arbi- trarily assigned to corrosion products in the calculations, making some 138 allowance for fission products immobilized in cracks and crevices in the moderator, etc, 6.4 "Puel Cycle Optimization The designer has little control over some of the independent variables that affect the fuel cycle cost., For instance, the maximum allowable fuel temperature is fixed by necessity of limiting corrosion rates, For some variables the fuel cost‘may decrease monotonically as the variable tends toward an extreme value, but other costs may increase, For example,'deu creasipng the external volume of fuel salt decreases inventory charges but " increases pumping costs, A plausible and conservative external volume was selected for the reference design optimization; however,.efféct.on,fuel cycle cost of decreasing the external volume is easily estimated. . After the values of such fixed or limiting variables were established, there remained several which required optimization simultaneously with re- spect to the fuel costs. These variables were désignated the "key" variables, ] | The key independent variables were found to be the diameter (D) of the core, the volume fraction (F) of fuel in the core, the concentration (M) of thorium in the fuel salt, and the processing rate (R). The second and third combine to fix an important subsidiafy variable, the C/Th atom ratio (an indication of the degree of moderation of the system). Fixing all four and then satisfying the criticality equation together with the equilibrium isbtope equation results in fixing the Th/U ratio, breeding ratio, fuel cost, etc. | | - Exploratory calculations showed that the fuel cycle coét is a rather sensitive function of the C/Th ratio, but is insensitive to the diameter of the core in the range from 15 to 20 feet. Relative breeding ratios and fuel cycle costs for a series of calculations are shown in Table 6.4, On the basis of these results, the 20-foot core'having a fuel volume - fraction of 10 per cent and using a fuel salt containing 9 mole per cent ThFu was selected'fbr further study. The optimum processing rate for this wh 139 Table 6.4, Conversion Ratios and Fuel Cycle Costs 1000 Mwe Molten Salt Reference Design Reactor Processing at Rate of Two Cubic Feet per Day Case Core Diam Vol. Frac, Mole % C€/Th Th/U Conv. FCC ft. Fuel ThPu Ratio m/kwhr 15 0.18 13 107 10,6 0,91 1,09 17,7 0.18 13 107 12,4 0.96 1.02 20 0,18 13 107 13,5 0,99 1,04 15 0.10 ] 293 21,0 0,84 0,73 17.7 0,10 9 293 23,1 0,87 0,89 6% 20 0.10 g 293 24,0 0,90 0,70 5 5 5 N F W N | 15 0,107 468 20,0 0,68 0,82 8 17.7 0.107 468 21.7 0.72 0,78 9 20 0,107 468 22,9 0.7% 0,77 *Reference case, preferred over Case 5 because of lower power density, lower velocities, etc, combination of key variables was then determined from results listed in Table 6.5, | | The numbers given in Table s,slare plotted in Fig. 6.1. The fuel cost has a minimum somewhat belowfo 7 mills/kwhre at conversion ratios lying be- tween 0,85 and 0.9, Sllght changes in cost assumptlons, etc. could shift 'the location of this mlnlmum over a wide ‘range. To the left of the cost mlnlmum, loss of neutrons to fission products increases burnup costs more rapidly than processing costs decline; to the 'rlght the processzng losses outweigh the gain in conversxon. The extreme in the conversion ratio results from the fact that proce581ng losses in- crease linearly with process;ng rate whereas loss of neutrons to fission 'products decreases only 1nverselyo ‘The processxng costs used 1n computlng Table 6.5 were those associated ;w;th a central Fluoride Volatlllty Plant (Sec. 5.2) but followed, where appllcable, the prescription glven in the Guide (52). The "turn-around" time was 2 days for all of the cases listed. FUEL CYCLE COST, mills/kwhe 1.1 1.0 0.9 0.8 0.7 0.6 0.5 004 140 ORNL-DWG. 65-7912 O Central Fluoride Volatility Processing @ Reference Design Case CONVERSION RATIO Fig. 6.1. MSCR Fuel Cycle Cost Versus Conversion Ratio. & 141 Q_} . Table 6.5, Processing Variables Case Cycle Time Volume Rate Weight Rate Conv, Fuel Cycle Cost days £t3/day kg Th/day Ratio mills/kwhre 1 3000 - 0,83 27 0,835 0,71 2000 1.25 40 0.868 0.68 3% 1500 1.67 53 0,895 0.68 4 1250 . 2,00 64 0,904 0.70 5 1000 2,50 80 0,917 0.74 6 500 5,0 160 0,930 1,01 7 250 10,0 320 0.908 1,64 “Reference Design Case 6.5 Reference Design Reactor The MSCR is capable of producing ppwer at a fuel cycle cost, including salt charges, of 0.7 mills/kwhre at a conversion ratio of 0.9, The most important uncertainties in this calculation arise in connec- tion with (a) the behavior of xenon in the core, (b) costs estimated for the Fluoride Volatilityfprocsssingsplant, (c) validity of the base charges. assigned to the materials_and“(d) costsof-packaging the spent fuel for shipment. The influence of xenon behavior is examined in Sec. 6.6, and the -costiassuhptions in Sec, 6.7, If the losses to xenon assumed for the reference design case are attainable (and it should be pdssible to achieve ‘the assumed'performance by improfiing'the graphite and the sparging process), “then all the uncertalnty resides 1n the proce551ng cost, Since processing -costs are only about 0,08 mmlls/kwhre (Table 6. 10) even a large error would | _not s;gnlficantly influence the total fuel cycle cost.r_ | Uncertalntles in regard to technical feasibility arise in connection 'thh compatibility ‘and stability of the graphite moderator (Sec. 4,2.2), of the reactor vessel and its 1nternal structure (Sec. 4.2.2), ‘Also, the - ~i; | ~ hazards assoc1ated with the proposed methods of control (Sec. 4.2.3) have not been evaluated. 142 6.5.1 SEecifications Case 3 was selected for the reference design before the complete curve was generated; its fuel cost is not significantly greater than the minimum. The reactor characteristics and operating data for the reference design ere given in Section 4, nuclear data in Table 6.6. Teble 6.6. Nuclear Characteristics of 1000 Mwe Molten Salt Converter Reactor Case No- 3 Carbon/thorium atom ratio ~300 Thorium/fissile uranium atom ratio ~27 Fraction of fissions in thermal neutron group *0.82 Fraction of fissions in 233U . 0.15 Fraction of fissions in 33U 0.85 Ratio of total fissions to fissions in 1.0018 233\ ang 235y Mean eta of 233y 2.253 Mean eta of 23°U 1.979 Mean eta of all fissions¥* 2.219 Effective resonance integral of 232Th 66 barns Thermal cross section of ?23Pa 39 barns Effective resonance integral of 233Pg 900 barns Average power density in core 14.1 kw/liter Ratio peak-to-radial average power density 2.1 Maximum graphite exposure rate, nvt/yr Neutron energy >0.1 Mev 5.0 x 102! - Neutron energy >1.0 Mev 1.8 X 10%% Average thermal flux 3.7 x 103 Average fast £lux 4.2 x 1013 Fissions per 2357 atom added 9.2 Fissions per fissile atom processed 1.5 Exposure, Mwd/tonne of thorium 47,000 Specific power, Mwt/kg fissile 0.9 | *Neutrons produced from all sources per absorption in and 235U. 233U ey 143 6.5.,2 Neutron'Eceheh§ » From Table 6,7, it is seen that flSSlonS in 232Th contribute very little to neutron productlon. The fast fission cross sections of thorium are appreciably less than those of 238y, moreover, the fast flux in the MSCR is not particularly high (Table 6.6), Also, the thorium is rather dilute compared to concentrations customarily proposed for blankets. The nuclear loss resulting from absorptions .in Pa is appreciable but not serious. It could be reduced by reducing further the volume fraction of fuel in the core, but this is already about as low (0.1) as seems tech- nically feasible; a further decrease would probably add more to the fuel cost in terms of increased inventory charges (since concentration of tho- rium and uranium in the fuel stream would increase, while the external volume would remain the same) than would be saved in terms of fuel replace- ment costs. Increasing the external volume in order to dilute the Pa would not be economical for the same reason. | The nuclear loss to Pa could be reduced by removing the Pa rapidly from the circulating stream and holding it until it decays to 233y, In order to be effectlve, such a process would have to treat the entire fuel stream for Pa removal in a period not greater than ten days (mean life of Pa is about 40 days). Thus an extremely simple and efficient process is required, as for example, the passage of the fuel stream through beds where Pa is selectively absorbed and retained until 233y is formed, which then desorbs in the preaence of large amounts of uranium in solution. No such process is presently known, although some work has been reported (19 p. 117, 74) on the prec1p1tatlon of protactlnlum oxide from molten salt solutions contaxnlng up to 2000 ppm of uranlum by contactlng the melt w1th BeO, Tho2 or 002 apprec;ably and add 0. 01 units to the conversmcn ratlo. Development of such a process might reduce the losses to Pa For reasons lndzcated in Sec. 6, 1, the concentratlon of 23“0 used 1n the‘equlllbrzum calculatxenrwas,averaged over a period of 30 years. Although the reactor was assumed teibe initially fueled with enriched ura- nlum containing 1.0% 234y and to be supplied with make-up fuel of the same vcomposxtlon ninety-nine per cent of the 234y is formed by transmutatlon of 233y, 1t disappears from the reactor by transmutation to 235U and loss in 144 Table 6,7. Neutron Economy in the 1000 Mwe Molten Salt Converter Reference Des1gn Reactor Item Captures Fissions Absorptions. 232Th 0.8535 0.0011 0.8546 233py '0,0084 0,0000 0.0084 233y 0.0888 0,7488 0.8376 234y 0.0572 0.0002 0,0574 235y 0.0301 0.1323 0.1624 236y 0.0184 - 0,0001 0.0185 237Np 0,0074 - . 0,0074 238y 0,0029 10,0000 0.0029 Carrier salt 0.0387 - 0.0387 Graphite 0.0564 - 0.0564 135%e 0.0170° - 0.0170¢ Other fission products 0.0867 .- 0.0867 Corrosion products 0.0082 - 0.0082 . Delayed neutrons 0.0046 - 0.0046 Leakage 0,0513 - 0.0513 Neutron yielda 2,2121 Processing lossesb 0.005 Net conversion ratiod 0.90 ®Neutrons per neutron absorbed in 233y ana 235y, bProcesszng loss of 0, 3%/pass for 233y and 2350 and undecayed 2 33pa. “Loss corresponding to graphlge having gas_porosity of 1% and diffusion coefficient of 1l0™° compared to current graph- ite properties of 10% and 1077, ‘ dExcludmg captures in 238y and correctlng for fissions of thorium, O ¥ 4 * 145 the processing cycle (0.3% per pass). Since 23filis inferior with respect to 233y as a nuclear fuel, it is advantageous to keep the concentration of 2347 as low as possible. However, the designer has little control over this, inasmuch as the only menas of removing it is to sell spent fuel to the AECQ and this, as shown in Sec. 6.8.3, is not advantageous., The concentration of 236U used was also averaged over a period of thirty years, starting with a clean reactor containing no 236y, 1In this case, the only source is capture of neutrons in 235y; the sinks are trans- mutation to 237U (which decays promptly to 237Np) and losses (0,3% per pass) in the processing cycle. The only effective ‘control the designer has 6ver?2360 is by varying the conversion ratio thus varying the amount of 235y fed to the system, and by sale of spent fuel, It is assumed that 237Np was removed 100 per cent per pass through the processing cycle. Parasitic captures are appreciable (0,007% units on the conversion ratio); however, special processing for this reason does not appear to be worthwhile, although it might be for other reasons. Parasitic captures in carrier salt resulted in a conversion loss of 0.039 units. Of this, captures in ®Li contributed 0,014 units, The grade of salt used (99,995% 'Li) appears to be about the best available at attractive pfices. On the other hand, use of inferior grades would not result in lower fuel costs. . The best way to control losses to ®Li is to recycle carrier salt from processing'instead of discarding it as was assumed in the reference design. The p0851b111t1es are examlned in Sec. 6.8, where it is shown that about 0.01 units mlght plaus;bly be saved on the conversion ratio, and about 0,08 mills in replacement costs, ~~ The fuel cost was 0pt1m1zed wmth reSpect to para31t1c captures in hgderator and neutron ;eakagersxmultaneously as descr;bed in Sec, 6., Losses to graphité might be decreased by decreasing the C/Th ratio, but the | gain fiould be more thahz¢ff$et by losses in eta of 233U and increased leakage, Leakage decreases slowly with increasing diameter, but fuel cost is insénsitive aé'shown'in'Téble 6.4, Efforts to reduce the leakage by use of a graphlte reflector were not successful (Appendix M), largely because of the necessary presence of a fuel annulus at least one inch thick . 146 o v <. F at the periphery of the core as a result of tolerance allowance for dif- - ferential thermal expansion, - The estimate of the loss of neutrons to kenon (0.017 units on C.R,) was based on an assumed diffusion coefficient of 107° em?/sec., a porosity accessible to gas of 1.0 per cent, and a sparging rate of 10 per cent (16tft3/sec) of the circulating fuel stream with 100 per cent removal of ¥enon per pass (Sec. 6.8.2). The prospects are good for the development of ~grades of graphite that would reduce the lossesito xenon in the MSCR to no more than 0,005 units on the conversion ratio. | Captures in samarium (1*9Sm and 151Sm) result in a loss of 0,013 units in conversion ratio. This loss is independént of the processing rate ex- cept at very short cycle times of the order of days. 'Thus,_tfiére_is'not much prospect of reducing this loss except by the application of some simple, rapid process similar to that suggested for 233p3 above. Captures in other fission products can be controlled by varying the processing rate. The savings from greater fuel conversion must be balanced against increased.processing'coSt. For the particular price structure and nuclear properties assumed, the optimum rate of processing is 1,7 ft3/day (53 kg of Th/day). The corresponding loss in conversion ratio due to captures in other fission products is 0,074 units. In this calculation, it was assumed that xenon is sparged as described above, noble metals are reduced by chromium in metal structures to the zero valence state and "plated out," énd all other fission products are removed by passage through the processing cycle which, at the costs estimated in Sec. 5.1, optimized . - at a cycle time of ~1500 days. | | If the procéssing cost schedule (Sec, 3,2 and 5.1) were to change in the direction of lower costs, the optimum processing rate would increase, and the fission firoduct.captures could be decreased. Although improve- ments and economies in the Fluoride Volatility process are to be expected, the remote opefations ahd maintenance costs of this process are likely to remain high; therefore the process eventually used should be as simple as possible, The solution may lie in the direction of distillation or frac- tional crystallization (80, p. 80), or perhaps extraction with liquid — metals, L)L 147 The estimated loss in conversion ratio to corrosion products was 0,006 units; a loss of 0,008 units was allowed in the calculations (Sec. 6.3). Since the concentration of corrosion products (Fe, Ni, and Cr) in the melt appears to reach an equilibrium in times that are short compared to the processing cycle time, it seems unlikely that the processing will have much influence on the cohcentratibns of corrosion products, and the associated loss of neutrons seems unavoidable, 6.5.3 Inventories and ProcessingfiRates It may be inferred from Table 6.8, Column 1, that the specific power of the MSCR is 0,35 Mwe/kg fissile, which is comparable with many of the advanced systems currently being put forward. The exposure (about 40,000 Mwdt/tonne of thorium) is of the same order as that of "competitive" | reactors. Table 6.8. Inventories of Nuclear Materials = 1000 Mwe Molten Salt Converter Reference Design Reactor Inventories, kg Material Reactor o Plant Processing Total 232p, 80,000 6500 86,500 233p,y 95.5 7.8 103 233y 2110 172 2,282 234%y - . ngu 39,6 524 235y . w20 ' .us% 465 236, | 682 56 | 738 237xp 49 5 - 54 238y 109 11 120 ©Salt . 109,000 8,900 . 117,900 Fission products 4,800 360 - 4,760 . .Corrosion products 82 - . 7. . 88 , *Including 10,6 kg in reserve to keep reactor in operation for 30 days after unscheduled lnterruptlon of recycle from proce581ng planto ' - 148 The processing rates given .in Table 6.9 are the amounts vemoved daily from the circulating fuel stream. These daily increments are accu- mulated for 116 days to form a pfocessing batch. The average fission pro- duct activity in the material as processed corresponds to an effectlve holdlng time of about 90 days (based on the Way-Wigner correlation as re- ported in reference 55, p. 81). Thus, a process batch contains 6500 kg of thoriumj this is processed at the rate of 1000 kg/day for uranium recovery, taking &7 days to process and two days for "turn-around." Table 6,9. Process and Make-up Rates = 1000 Mwe Molten Salt Converter Reference Design Reactor | \ - Rate, kg/day Material To . Processing Make=-up 232n 53.3 3.56 233Pa 0.064 23“0 1.400 U 0.322 0,003 23 0.280 0,324 237U 0,463 >aaP 0.033 U 0.073 Salt | 72.7 72,7 Fission products 2,94 Corrosion products 0.055 6.5.,4 Fuel Cycle Cost The fuel cycle cost for the reference design reactor, which is ngér optimum on the bases chosen, comes to 0,68 mills/kwhre. A breakdown is given in Table 6,10, A ‘ Inventory of fissile materials costs about 1/4 mills/kwhre, when optimized with respect to the processing rate. It could possibly be re—,'- duced at a given processing rate by improving the graphite to reduce xenon ‘poisoning and by reducing the volume of fuel in the external heat transfer circuit at the expense of greater pumping power costs, O O ) 4 + sion ratio, Increasing this from 0.90, as in the reference design reactor, cessing) with the follow;ng mlnor exceptlons- "These, however, are oonSLdered wmthln ‘the reach of current technology, i.e., ' no developmental break-throughs are requlred. 'design of the reactor vessel and its internal members, it appears that " these can be solved. The problem in relafion to graphite is to produce 149 Table 6,10, - Fuel Cycle Cost = 1000 Mwe Molten Salt Converter Reference Design Reactor with Salt Discard Charges,* mills/kwhre Material Inventory Replacement Processing Total 2321y | 0,03 0.04 233pgy 0,01 233y | 10,18 - 0.08 | 235y L 0,04 0.16 | Total 0.26 0.20 0,08 0.5u4 Salt costs 0.06 0.08 . 0,14 Total charges - : 0.68 *Cost bases are given in Sec. 3 Replacement costs for fissile material are most sensitive to conver- to 0,95 would eht the cost in half, The increase in conversion ratio might be achieved by means discussed in Sec. 6.7.3. | The inventory charge for salt is strictly a function of the fuel salt volume, The replacement cost depends on the processing rate, and was optimized. A major improvemenf here would consist of adding equipment in the processing plant for reco#ering lithium and beryllium, The fuel cycleocoét of 0.68 mills/kwhre eStimated.forfthe MSCR is con- servatlvely based on the scale-up of proven technology (lncludlng the pro- 1. Technology of reactor vessel de51gn.‘ 2. Graphite technologye '3, Xenon sparglng technology Although difficult problems in gamma heating may be encountered in the 150 pieces of the size required that are chemically compatible with fuel salt and have porosity and permeability suitably:low with respect to xenon ab- sorption (Sec. 6.8.2). | . - | 6.6 Parameter Studies In this section, the effect on the fuel cycle cost of various assump- tions concerning the processing cost, and of several modes of processing are considered. 6.6.1 Processing Cost as 'Parameter The fuel cycle cost reported in Sec. 6.7 (Table 6.10) was based on the assumed use of a central Fluoride Volatiiity facility requiring approxi- mately $20 million in capital investment and having an annual operating ex- pense of $2 million (Table 5.7). In Sec. 5.4, this estimate was compared with other current estimates for similar plants and for Thorex'plants-of comparable capacity,’ In this section, the processing plant capital invest- ment and the operating cost are considered as'parameters, without regard to the kind of plant., The effect on the fuel cycle cost of the MSCR at various processing rates of varying processing costs for a plant having a nominal capacity of 1000 kg Th/day (30 ft3/day of reference design salt) | was calculated, 1In éll cases, the output of the reactor was accumulated until a processing batch of‘abbfit 6000 kg of Th was collecfed; this-Was . then shipped to the processing plant and processed at a rate of 1000 kg Th/day. The turn-around time was assumed to be 2 days. | The results of the calculation are presented in Fig. 6.2. The curves corresponding to daily charges of $20,000 to $60,000. It is.séen that; in the conversion ratio range from 0.8 to 0.9, the fuel cycle cost is not very sensitive to the processing cost; in the reference design, the fuel cycle cost does not exceed 0,85 mills/kwhre at a daily charge of $40 thousand, and only slightly exceeds 1.0 mills/kwhre for a charge of $60 thousand. At $u0 thousand,.the minimum fuel'cycle cosf_is less than 0.8 mills/kwhre at a conversion ratio of about-o.es; and the cost remains below 1,0 mills/kwhre for conversion ratios up to about 0,92, FUEL CYCLE COST, mills/kwhe 0 | _ 360,000 483 flq________,—aa’ .8 151 ORNL-DWG. 65-7913 | | Ty cpe¥e -~ CONVERSION RATIO Fig. 6.2. MSCR Processing Cost Versus Conversion Ratio. $40,000 — ) o 0.80 0.85 0.90 0.95 152 It was concluded that, on any reasonable cost basis, the optimum fuel cycle cost for the MSCR will not exceed 0.8 mills/kwhre, and that conver- sion ratios up to 0,92 can be obtained at fuel cycle costs not exceeding 1,0 mills/kwhre, 6.6.2 Effect of Xenon Removal The solubility of xenon in LiF-BeF2 is very low (107); in the refer- ence design reactor the equilibrium pressure is about 0,06 atmospheres. Xenon thus tends to leave the salt at any phase boundary. It can be re- moved rapidly by spraying a portion of the circulating stream into a space filled with helium or by subsurface sparging. It may form microbubbles , - clinging to the surface of the graphite moderator, and it will tend to diffuse into pores in the graphite; including pores inaccessible to the . salt, Xenon is also removed from the system by decay to 135¢s and by reaction with neutrons to form l35Xe, which is stable and has a low neutron capture cross section. ‘ Xenon poisoning in the MSCR was calculated by the method of Watéon and Evans (107), as shown in Appendix G, The important physical properties are the porosity, e, of the graphite (fraction of graphite volume accessible to xenon) and the diffusivity of xenon, D (cm?/sec). The key variables are the diameter of the graphite logs, the fuel circulation rate, and the sparging fraction, r, (fraction of circulating stream sparged or sprayed to removed xenon)., In the reference deSLgn, the logs are six inches in dlam- eter, and the circulation rate is 160 ftalsecoll_ In the reference design reactor, a gas-accessxble porosity of 0,01 and a diffusion coefflclent of 10~ & would result in a tolerable_xenon poison fraction of 0.017 neutrons per with a sparg;ng_fractlon of 0,1 (16 ft?/sec), : neutron absorbed by fissile atoms. These physical property values are an order of magnitude smaller than those of currently available graphite where e=0,1and D= 10" able in small pieces of graphite (107) and the control of xenon poisoning s however, the assumed values are both presently attain- in the MSCR appears to lie within the reach of developing technology. - The conversion ratio increases and the fuel cycle cost decreases with —~ decreasing xenon poisoning. Table 6.1l compares the calculated results for &/ 153 Table 6.11, Effect of Graphite Properties and Sparge Rate on Performance of 1000 Mwe Molten Salt Converter Reactor ............... Conversion Fuel Cycle Cost Case Xe P°rf ________ Ratio : mills/kwhre a 0,054 . 0.84 . 0.79 b 0,045 | (0.86)% (0,77)% c 0,017 0,90 0.68 4 0,001 . . ... 0,92 0.66 *Interpolated three cases: (a)fi"Worst" case, with a very porous graphite (say AGOT) and with no spargipg;'(b)'available graphite with eD = 107° and r = 0.1; (c) Reference Design Reactor with eD = 107° and v = 0,13 (d) "impermeable" graphite with eD = 0 and r = 0.1, It is seen that while avallable graphite (Case b) is not 31gn1f1cantly better than the "worst“ graphlte (Case a), nevertheless, the fuel cycle cost is only 0.1 mlll/kwhre hlgher than for "impermeable" graphite (Case d). It is concluded that the fuel cost is not very sensitive to xenon poison- ing, that it will be less than 0.8 mills/kwhre,'with available graphite, ‘and that with modest'imprOVements’over available graphite (to e of 0,01 and D of 10'5), the fuel”cost will be not more_tfiéfi“5;7 mills/kwhre., - 6. 6 3 Effect of Product Sale Wlthout Recycle At least two beneflts accrue from: the sale of recovered f1551le iso- ,.topes (as- UFg) to the AEC: (a) Make-up fuel (235U) would then be non- 'radioactive and could be compounded with fresh salt in a dxrectly main- _tained and operated facllity, (b) the reactor would tend to be purged of 236y, The second benefxt is really illusory, inasmuch as ‘the 236U produced "'Wlll eventually capture a neutron 1n some reactor somewhere, and therefore reduces ‘the value of the recovered 1sotopes by an approprlate amount. On the other hand, a penalty is incurred in that 2330 a superior Ffuel in thermal reactors, is also lost from the system. As seen in Table.6,12, the penalties outweigh the advantages considerably. 154 Table 6:12. Effect of Sale of Spent Fuel on MSCR Performance Case o - A | B Spent fuel is . . . Recycled : Sold Absorptions in 236 0.0185 ~ 0,0105 Mean Eta 2.21 | 2,11 Conversion Ratio h 0,90 0.82 Fuel Cycle Cbst, mills/kwhre Inventory Charges 0.32 0.36 Replacement Charges | 0.28 '1.38 Processing Charges 0.08 0.08 Production Credit ‘ - - °°53- Net Fuel Cost, mills/kwhre 0.68 1,29 6,7 Alternative Design and Cost Bases 6.7.1 Thorex Processing Cost Estimates The preferred method of processing MSCR fuel is by fluorination (Sec, 5.2), mainly because the processing-can‘conveniently be integrated with the reactor and thus achieve very low fuel cycle costs. However, for one reason or another, it may be desirable to process the spent fuel from the first MSCR installations in a central Thorek facility. Aécqrdingly, the facility specified in the Guide to Nuclear Power Cost Evaluation (52, 104) was modified appropriately as described in Sec. 5.3 to handle MSCR fuel, An allowance of $500,000 additional capital cost for the head-end treatment was made. Costs were calculated on the bases given in the Guide. The turn-around time was eight days and the shipping charge was $17/kg for round trip. The assumed loss of fissile material was 1.3 per cent/péss. The fuel cycle costs for the reference design reactor'(Tabies 6.6 through 6.10) were recomputed using a cost schedule estimated from the Guide (52), as shown in Table 6.13, where they are compared to correspond- ing results for central Floride Volatility processing (Sec. 5.2). 155 Table 6 13, Effect of Processing Method ------- on ‘Fuel Cycle Cost Central Central Fluoride Thorex Volatility Proeessing cost; $/kg Th 75.2 uy,8 Processing iosses, per cent/pass 1.3 0.3 Conversion Ratio 0.89 0.90 Fuel cyecle cost, mills/kwh Inventory charges 0.33 0,32 . rRepieCemenf'charges' ' 0.24 0.28 Processing cherges 0,17 0.08 : | Tctal mllls/kwh -::?;: - 0,68 ................. Thus, even though processed through the AEC reference plant, the costs of which appear to be conservatlvely high, the fuel cycle cost for the ref- erence design MSCR will not exceed 0,75 mills/kwh., 6.7.2 'Reactor-Inteégrated Fluoride Volatility Processing Several advantages can be realized by integrating the processing with reactor operations. The principal saving results from sharing reactor shielding and remote maintenance equipment. Savings in laboratory facil- % ~ ities and personnel are alse~importantc Shipping'coets and associated re- ;;ce1v1ng fac111t1es are elxmznated v A 30 ft /day central fluorzde volatlllty plant requlres a capltal out- 'lay of about $25 million and would need ‘to service flfteen to twenty 1000 Mwe MSCR's in order to achieve the unit processing costs estlmatedrin SectionVS 2. An'integrated plant requires a much'smallerrinvestment and the unit processxng cost will be, of course, lndependent of the number of ~ reactors in use. ) An'integrated facilityVShculdrbe designed for continuoue flow process: ing, at least in the fluorinator and the UF, reduction reactor, in order \EJ that the equipment and the volume of fuel-salt held up might both be small, 4 ..A_um..-‘..._... 156 6.7.3 Reactor-Integrated'Precipitafion'Process At a processing‘rafe of 1.7 ftalday, the concentration of rare earth fluorides in the fuel salt will be approximately 0.5 mole per cent, which is perhaps slightly in excess of the solubility limit at the minimum fuel temperature of 1100°F. This suggests that it may be possible to remove rare earths fission products from the MSCR fuel stream by fractional crys- tallization — an e#tremélywattractive possibility, for such a process could conveniently be carried out in the reactor cell and closely integrated with the reactor system. The steps involved are exceedingly simple, involving only the transfer of liquids and heat, and is therefore'inherently safe and' economical, , , , The fission product neutron poisoning estimated for thé referénce de~ sign can be approximately matched by charging every day five cubic feet of fuel salt containing 0.5 mole per cent rare earth thorides to a crystal- lizer, The salt is cooled to 900°F, which is 13°F above the temperature at which solid solutions containing Th or U separate. At this'temperature, the solubility of the rare earth fluqrides is 0,2 mole per cent or less, judging from fhe data of Ward et al., (108, 106). Thus about 60 per cent of the rare earths will precipitate or "freeze" on the walls of the crys- tallizer. The total mass of the solids will be only 6-8 kg, After the fuel salt is returned to the reactor system, the fission products are dis- solved in flush salt \ The process deseribed should effectively remove rare earths from the fuel salt, It will not remove alkali metals, alkaline earfhs, and mis- - cellaneous other metals. These will accumulate in the fuel, but their in- . growth can be partially compensated by operating the freeze-process at a slightly more rapid rate and maintaining the rare earth concentration at say 0.45 mole per cent or by discarding barren fuel-salt in a 1500 day cycle, | | 6.8 Evolution of a Self-Sustaining MSCR Although the MSCR concept may not have the capability of evolving into a breeder reactor (see Sec. 7.1) having a doubling time less than 25 years, » v » 157 ‘the possibility exists, however, that it could; without increase in power costs, achieve a net conversion ratio slightly greater than unity, and thus become self-sustaining and independent of outside supplies of fissile isotopes. Conditions under which this might be achieved are listed and discussed below, 6.8.1 Reduction of Leakage With the advent of separated ?2Mo, it becomes feasible to surround the core of the MSCR with a thin blanket of high-density thorium salt (25% ThF, 75% LiF). This salt, having a liquidus temperature below 1100°F, could be circulated slowly through 6-inch diameter molybdenum tubes replacing the outer two layers of grephite'moderator legs, and replacing 1l-ft end sec- tions of the central logs. The isotope I32Mo is a magic nuclide, having a 2200 m/s cross section of 6 millibarns. The epithermal cross section is currently being measured athRNL, and preliminary results indicate that the resonance integral is also:very small, Thus, structural molybdenum should capture only a negiigible_fraction'of neutrons. It would be necessary:to:remove the bred 233y papidly from the fertile stream for two reasons: (a) fissions in the fertile stream would tend to increase the leakage, (b) fission products in the fertile stream would capture neutrons, If fhe fertile stream were processed rapidly by fluori- nation, the concentration of 233U could be kept very low, fissions would be suppressed, the inventory charge for 233U would be largely avoided, . The fertile stream carrier'salt could be recycled without furteer treatment. The rapidity of the preCeSSing; however; implies the use of an on-site, reactor-lntegrated faczllty such as that described in ‘Sec. 6.9,2, By this means perhaps half the leakage neutrons could be saved, addlng ~_about 0.025 units to the converSLOn ratio. -1 # 6. 8 2 Reduction of Xenon Captures _r " In the reference desxgn, graphlte propertles an order of magnltude - better than those characterlstlc of current graphites were assumed, result- ing in a loss ih conversion ratio due to captures'in 13_5Xe'absorbed in the graphite of only 0.017 units. By further improvements (e.g., by spraying a thin coating of 92M6 on the moderator logs and carburizing to prevent xenon 158 — from penetrating the moderator) perhaps another 0,01 units on the con- version ratio could be saved., 6.8.3 Reduction of Fission Product Poisoning Somewhat over 0,085 units weré lost from the conversion ratio as a result of captures in fission products in the reference design (Table 6.7). The processing cost in a -central Fluoride Volatility facility w&s about 0.08 mills/kwhr., If an on-site, reactor-integrated Fluoride Volatility and - HF-Solution facility were used (Sec. 2;2.lu),_fhe rate of removal of rare earths could be increased perhaps by a factor of 10. Solubles (Cs, Ba, etc.) could be purged by discarding salt in a 1500 day cycle, as in the ' . reference design. If this fiererdone, the loss of conversion ratio to fission products could be reduced to about 0,010 units, - 6.8.4 Imprdvement'of‘Mean'Eta'and'Reduction of 236y captures Other than by varying the C/Th ratio, the designer has no direct con- trol over these., Nevertheless, the above improvements in neutron economy have an effect on the conversion ratio thét is greater than fheir cumulaf' tive sum, for 233y is superior to 235U in respect to neutron production, and moreover yields a fertile isotope (23"0) upon capture of a neutron, An increase in relative concentration of 233U by any means increases the number of neutrons available for breeding, and reduction of 235U feed rate results in a decrease in 236U concentration. o _ The above listed improvements should result in an increase in eta of - | 0.01 units and a reduction of captures in 236y b§0,01 units, at least.’ 6.8.5 ’Ultimate'Bréé&ixll:g:"é;{;fié;itéf'MSCR The conversion ratio in the reference design is about 0,90, With the improvements listed above, conversions slightly in excess of 1.0 may be . achieved, as shown in Table 6.1k, S R Thus, the MSCR may be capable of evolving stepwise into an economical, self-sustaining breeder reaétor with fuel cycle costs probably in the range of 0,7 - 1,0 mills/kwhre, .' - o~ It should be emphasized that the limiting conversion ratio estimated A for the MSCR does not apply to molten salt breeder reactors., It has been o . » A ‘ j - 159 Table 6,14, Ultimate Breeding Potential of Molten Salt Converter Concept Conversion ratio in reference design 0.90 Savings due to: Reduction in leakage 0.025 Reduction in xenon captures 0.010 Reduction of fission préduct captures : 0.075 Improvement in eta and reduction of captures in 236y 0.020 Ultimate conversion ratio 1.03 shown (3) that two-region, twé-fluid, thermal reactors optimized with respect to breeding are capable'bf achieving doubling times of 25 years or less, (See also Sec. 1.7.) In addition, the advent.of structural 92Mo makes pOSSlble, in prmnc1ple _the design of two—reglon, two~fluid, fast molten-salt reactors which may have doubling times as short as ten years, which will not need to use separated 6Li in the ‘carrier salt, and which therefore may be processed econommcally by ‘fluorination only. 160 7. MSCR CAPITAL INVESTMENT, FIXED CHARGES, AND OPERATING EXPENSE 7.1 Introduction ' The equipment, auxiliaries, and auxiliary services described in Sec. 4 were costed for ORNL by Sargent and Lundy Engineers of Chicago, Illinois (95, 96). Equipment arrangement drawings sufficient for piping take-offs and building cost estimates were made. Details of these studies are given in the referenced reports. 7.2 Surmary of MSCR Capital Investment In Table 7.1 are listed the principal items of cost in the MSCR ref- erence design. A detailed breakdown is given in Appendix N. The account numbers correspond, where applicable, to the AEC systems of accounts given in The Guide to Nuclear Power Cost Evaluation (52)._ Table 7.1, 1000 Mwe Molten Salt Converter .. ..Reactor.Capital Investment Fission energy release rate, Mwt | 2500 Mwt Net station power - | 1038 Mwe Gross station power 1083 Mwe Station efficiency ' 41,5% Heat rate 8220 Btu/kwhr Plant factor , 0.8 Total capital investment | $143/kwe Direct Construction Costs 21 Structures and Improvements 211 Improvements to site $ 501,500 212 Buildings | . 5,465,450 218 Stacks , 31,000 Reactor Container ' (Included in 212) Total Account 21 (5,997,950) 4 161 Table 7. l. Continued 22 Reactor Plant 221 222 223 224 225 226 227 228 229 Reactor equipment Heat transfer system Fuel fabrication and handling system Fuel processing system waste disposal Low-level radioactive waste disposal Instrumentation and controls Feed water supply. Steam, condensate, and water piping Other reactor equipment Total Account 22 23 Energy Conversion System 231 232 233 234 235 236 237 238 Turbo-generator unit Circulating water system Condensers and auxiliaries Central lubrication system ‘Turbine plant instruments & controls Turbine plant piping Auxiliary equipment for generators Other equipment Total Account 23 24 Accessory Electrlcal Equipment 241 242 243 244 245 246 247 Swmtchgear Switchboards Protection equipment Electrical structures Conduit | Power and control wiring Station service equipment -Total Account 24 25 'Mlscellaneous Plant Bqulpment 251 252 253 Cranes and hoists Air compressors and vacuum pumps Other Total Account 25 Total Diféct Cohstruction.cost Indlrect Constructlon Costs E " Construction overhead (20% of direct labor) General and administration (2.5% of direct costs) Subtotal $ 8,823,300 23,609,700 1,517,200 (Not included) 361,150 1,100,000 4,939,500 7,925,000 3,048,500 (51,324,350) 21,495,000 1,644,200 3,104,900 36,000 426,000 (Included in 228) 137,000 (Included in 228) 26,843,700 637,400 286,000 131,600 213,200 210,200 2,281,900 615,000 4,375,300 195,000 64,900 540,000 - 799,900 89 341 200 2,333,300 5,775,500 (97,450,000) 162 Table 7.1l. Continued ............. Miscellaneous costs (1.,2% of subtotal) Subtotal Engineering D2sign and Inspection Architect-engineers (11.1% of subtotal) - Subtotal Nuclear engineers (3.8% of subtotal) Start-up expense (35% annual O&M expense) Land and land rights - Subtotal Contingency (10% of subtotal) Subtotal Interest duripgfcdhstruétion (9.4% of subtotal) Total Indirect Construction Costs Total Construction Costs Intermediate Coolant-Salt Inventory® TOTAL MSCR CAPITAL INVESTMENT $ 974,500 (98,424,500) 10,925,000 (109,349,500) 4,155,300 746 ,900 360,000 (114,611,700) 11,461,200 (126,072,900) 11,850,900 48,582,600 137,923,800 10,951,800 $ 148,875,600 *Including interest during startup 7.3 MSCR Fixed Charges The fixed charges were computed in accordance with the instructions in The Guide to Nuclear Power Cost Evaluation, Vol. 5, Production Costs, Tables 7.2 and 7.3. page 510-2 (52), for an investor-owned public utility, and are shown in " ) ¥ 163 Table 7.2, Nation~Wide.Approximated Fixed Charge Rates Percent'Per Year Depfeciating: - Non-Depreciating Profit on investment ' 6,75 6.75 Depreciation (30-yr sinking fund) 1.11 -- Interim replacements 0.35 - - Property insurance 0,40 0.40 Federal income taxes - 3.40 | 3.40 State and local taxes - 2.45 : 2,45 ‘Total 14,46 13.00 Table 7.3. 1000 Mwe Molten Salt Converter Reactor Fixed Charges Rate Annual Expense Power Cost Item Investment %/yr $/yr Mills/kwhre Depreciating capital 137,600,000 14,4 19,950,000 2.74 Non-depreciating capital ";i_ o . ‘Land, ete. . .- 360,000 13.0| Coolant 10,950,000 13,09 1,530,000 0,21 Working capital -~ 450,000 13,0] | Nuclear insurance @~ ==== 340,000 0.05 Annual fixed charges 21,820,000 3,0 164 7.4 MSCR Operating and ‘Maintenance Cost Estimate The MSCR is a szngle reactor, single turbzne plant and, although its power generatlon capacity is hlgh its manpower requirements are relatively low because of the single unit operation. Plant personnel totals 101 with a cost of $872,000 per year, Materials cost $220,000, Maintenance, in- cluding provision for periodic equipment overhafil, special services pro- vided by off-site personnel and qrganizations,‘totals $800,000, With an allowance of 14 percent for central office expense, the total cost is $2,154,000 or 0,30 m/kwh. ' “The ‘Guide to Nuclear Power Cost Evaluation (52) criteria were followed where applicable in determining plant organization. 7.4.,1 Labor and Materials { The manpower requirements of the 1000-Mwe MSCR plant are shown in Table 7.4 for operétions and general supervision. Routine operation of the Plant may require fewer people, particularly on the technical staff, Re- view of reactor plant personnel requirements developed by Sargent and Lundy (94), and Kaiser Engineers (51) for 300 Mwe (net) plants are compabed in Table 7.4 for single unit systems. -In practice, the actual distribution of manpower may shift, but the total labor cost should remain approximately as shown. For example, one storekeeper may be insufficient, in which case an engineering assistant or maintenance mechanic helper might be repléced by a stores clerk. In general, the turbine room operation includes, in addition to the turbo-generator proper, (a) water supply and disposal systems, e.g., sani- tary service, treated and circulating water systems, (b) boiler feed systems which include boiler feed pumps, steam clrculators, and Loeffler boilers, and (c) the turbine auxiliaries, e_,_go lubrlcating oil systems, liquid and gas coolant systems, and instrument and compressed air systems, The reactor operation includes (a) salt charging systems, (b) salt with- drawal systems, (c¢) salt shipping facilities, (d) liquid and gaseous waste disposal systems, and (e) pressurized gas supply and treatment systems, These facilities are staffed on a semi-automated basis; for fully automated operatlon, the manpower requirements would be less, o~ i ) 165 Table 7.4. Personnel Requirement Estimates Sargent & Lundy Kaiser AEC Guide MSCR 150 Mw - 350 Mw 300 Mw 300 Mw(PWR) 1000 Mw Plant Management, 9 4, y 7 Office & Stores | Operating Dept. 47 32-39 36 49 Technical Staff - 26 6 6 16 Maintenance Dept. | 13 20 13 29 Total o 85 62-69 59 101 7.4.,2, Operation and Maintenance Cost The estimate of $2,154,000 shown in Table 7.5 does not take into account any unforeseen diffiéulties and expenses that may be encountered in the MSCR, It may well be that requirements for personnel and equipment for maintaining a radicactive molten salt system are greater than estimated - possibly by as much as a factor of three. This cannot be accurately determined until maxntenance procedures have been more clearly defined and the reactor plant desxgned 1n greater detail than was possible in the present study. Based on 1038 Mwe net and a plant factor of 0.8, the contribution to the power cost is 0,3 mills/kwhre. Table 7.5. 1000 Mwe MSCR Annual Operatlng | and Malntenance Expense Salary or Personnel Annual Wage Rate Required Expense Wages & Salarles Plant Management 7 : ___Statmon Supt., : 7'$ lS,OOO/yr 1 $ 15,000 - Ass't Supt, 12,000/yr 2 24,000 Clerk-Steno 2.50/hr 1 5,200 Clerk-Typist 2.31/hr 2 9,600 Clerk-Steno 2.50/hr 1 5,200 7 59,000 e o 166 Table 7.5. Continued Technical Staff Supv. Eng. (L)% Nuclear Eng. (L) Engineer | Health Physies Supv, Eng. Ass't. Lab Technician Radiation Protection Operating Staff Shift Supt. (L) Senior Control Oper. (L) Control Oper. (L) Turbine Oper. Equipment Attendant Special Operator (L) Janitor Watchman Maintenance Staff Maintenance Supt. Foreman Instrument Mechanic Electriecian Pipe Fitter-Welder Machinist Mechanic Helper Total Labor Fringe Benefits at 20% Total Wages £ Salaries Materials for Routine Operations 0il Supply Gas Supply Treated Water Coolant Salt Make-Up Office Supplies Laboratory Supplies & Chem *#(L) Denotes licensed reactor operator. Salary or Personnel Annual Wage Rate Required Expense 11,000 /yr 1 11,000 9,600/yr 1 9,600 8,400/yr 3 25,200 8,400/yr 1 8,400 6,000/yr 3 18,000 2,85/hr. 5 29,700 3.25/hr. 2 13,500 16 115,400 10,800/yr 5 54,000 3.75/hr 6 46,800 3.65/hr 4 30,400 3.50/hr 10 72,800 3.00/hr 8 50,000 3.50/hr 9 65,500 2.25/hr 2 9,400 2.25/hr 5 23,400 49 352,300 10,800/yr 1 10,800 7,500/yr 3 22,500 3.25/hr 6 40,600 3.25/hr 5 33,800 3.25/hr 2 13,500 3.25/hr 2 13,500 3,25/hr 8 54,100 2.65/hr 2 11,100 29 200,000 101 725,700 145,300 872,000 84,000 4,000 2,000 40,000 15,000 5,000 167 Table 7.5. Continued Salary or = Personnel Annual Wage Rate Required Expense Miscellaneous (e.g., radiation 50,000 protection, clothing & equip. Consulting Services 10,000 Subtotal 210,000 Contingend& 10,000 Total Materials 220,000 Maintenance Turbine & turbine auxiliaries 150,000 routine maintenance materials Reactor & reactor auxiliaries 300,000 routine maintenance materials Turbine 3-yr overhaul (prorata) 50,000 Turbine system auxiliaries overhaul 50,000 Reactor system overhaul 100,000 Reactor auxiliaries overhaul $0,000 Subtotal 700,000 Contingency 100,000 Total Maintenance 800,000 Central Office, General & Admin, Expenses at 14 Percent 262,000 Grand Total Operating & Maintenance Expense $2,154%,000 168 8, RESULTS AND CONCLUSIONS The cost of electric power is commonly resolved into three components: The fuel cost, the fixed charges, and operation and maintenance expense. 8.1 Fuel Cost As shown in Sec., 6, the fuel cycle cost in the MSCR rangés from 2 mills/kwhr (electrical) down to 0.7 depending on the conversion ratio desired and the method and/or cost of processing assumed. For present pur- poses, the minimum cost associated with the reference design reactor summa- rized in Table 6.6 was selected as representative. This reactor is "near- term" and predicated on the scale-up of current technology with the ex- ception of the modefaton graphite, which was an order of magnitude better than currently available graphite in respect to porosity'and permeability. The fuel cycle cost, using Fluoride Volatility processing in a central plant with discard of carrier salt and contained thorium and recycle of isotopes of uranium, was reported in Sec. 6.7.1l. Table 8.1 1000 Mwe Molten Salt Converter | . .Reactor Fuel Cycle Cost CItem U Mills/kwhr Inventories Fertile 0,03 - Fissile 0,23 - Salt ' 0.06 0,32 Replacement ~ Fertile 0,04 - Fissile (95% 235y) 0.16 - Salt : 0,08 0,28 Reprocessing 0.08 0,08 Total, mills/kwhr 0,68 O 169 Lo 8;2"Fixed‘Char§es The capital investment for the 1000 Mwe (1038 Mwe net) station was estimated by Sargent and Lundy, Engineers from information supplied by ORNL, as reported in Sec. 7, and summarized in Table 7.1l. The investment comprised $137,56%,000 for depreciating capital items, and 511,311,800 for non-depreciating items (coolant salt and land). Fuel salt fixed charges are included in fuel cycle cost. Working capital was estimated according to the prescription given in the Guide (52), and is shown in Table 8.2. Table 8.2, 1000 Mwe MSCR Working Capital 1, 2,7% of annual dperating labor and fuel costs: $ 200,000 2, 25% of annual maintenance and materials: 250,000 Total Working Capital. $ 450,000 Similarly, the nuclear hazard insurance premium was estimated by prescription at $3u40,000 per year, The fixed charges are collected in Table 8,3, Table 8.3, 1000 Mwe Molten Salt Converter A ‘ f Reactor Fixed Charges : o e Rate Annual Expense Power Cost Item ~ Investment o v $/yr Mills/kuhr . Depreciating capital 137,600,000 14,46 . 19,950,000 2,74 Non-depreciating capital 11,760,000 13.0 1,530,000 0.21 Nuclear ihsurance'j“ B o 340,000 0.05 Annual fixed charges - 21,820,000 3.0 170 8.3 Operation and Maintenance Expense This expense was estimated in Sec. 7.2 and amounted to $2,154,000 per year, of which $1,020,000 was for maintenance and materials and the rest was for labor, supervision, and management. At 1038 Mwe net, the contri- bution to the power cost is 0.3 mills/kwhr (electrical). 8.4 Cost of Power The three components of the power cost are assembled in Table 8.4, Table 8.4, Cost of Power in a 1000 Mwe Molten Salt Converter Reactor Item _ _ Mills/Kwhr Fuel cycle cost Fixed charges 3.0 Operation and maintenance | 0.3 Cost of power, mills/Kwhr 4.0 In regard to the fuel cycle cost, it was shown in Sec. 6.8 that this would not exceed 1.0 mill/kwhr even though the fuel were processed in a 1.0 tonne/day Thorex plant costing up to $60 thousand per day to operate, Sgch a general purpose plant could provide processing of fuel from thorium reactors having a total power capability of about 20,000 Mwe at an exposure of 50,000 Mwd/tonne., | The energy conversion system, accessoéy electrical equipment, and miscellaneous plant equipment (Items 23, 24, and.25 in Table 7.1) are con- ventional items, and the estimation of their cost appears to be relatively unambiguous. Items 21 and 22 are perhaps subject to considerable uncer- tainty. But their total is only about $57 million, whereas the contingency ST item is $11 million. However, the possibility exists that Item 21 was L ) " 171 grossly underestimated, or that the effect of radiation shielding require- ments on building costs was underestimated. Supposing Item 21 to be $18 million (factor of 3), and allowing another $5 million for the reactor plant (10%), the investment comes to $166/kw,»and the fixed charges to 3,5 mills/kwhr, as an upper limit, Provision for labor in Sec. 7.2 seems adequate, Materials and supplies accounted for about half the operation and maintenance costs; if this were doubled, the contribution to the cost of power would be about 0,5 mills/kwhr, Collecting these probable upper limits on the cost of power in the MSCR, the sum is 5.0 mills/kwhr., 8.5 Breeding Potential of the MSCR The nuclear capability of the reference design MSCR is summarized in Fig., 8.1. With central Fluoride Volatility processing, the minimum fuel cycle cost is about 0.7 mill/kwhr and the conversion ratio about 0.9 at a processing cycle time of 1500 days., Increasing the rate increases the con- ~ version ratio to perhaps as high as 0,93, but the fuel cost rises steeply. The use of an on-site reactor-integrated (inside the reactor cell) Fluoride Volatility facility would increase the fuel cost by only 0.1 mill/kwhr in the 1000 Mwe station, | The use of an on-site reactor-integrated precipitation process for removing rare earths might reduce the fuel cost below 0.5 mills/kwhr, By-takihg Advénfége 6f fiofential impro&eménts.in'heutrén'economy-(but : retaining the'eSSentiai features of the MSCR) an upper limit on the con- version ratio of 1,03 was estiméted; This was interpreted to mean that the MSCR is'capable pf'evolving'into a self-sustaining reactor requiring only thorium feed, Outside é&fifiqe.offissiie isotopes wofild,not'be needed. This limitation does nbtwapply to two-region breeders, which were rSthn previously to be capable of doubling times as short as 25 years.(l). 172 ORNL-LR-DWG 76712 0.9 0.8 0.7 o -+ — — — 1 = Ultimate Breeding Potential of MSCR-—' FUEL CYCLE COST, mills /kwhe 0.6 ' ~ Symbols O Central Fluoride Volatilily Processing O Full Xenon Poisoning 0O Xenon Exclusion 0.5 © AReference Design Case ' I | 0.4 ' 0.8 0.9 1.0 ! CONVERSION RATIO | Fig. 8.1. Nuclear Capability of 1000 Mwe Molten Salt Converter Reactor. o ~ ey " 173 8.6 ‘Conclusions The 1000 Mwe MSCR requires a capital investment of $143/kwhr, The fixed cherges are 3.0 mills/kwhr, the fuel cost is 0.7 mills/kwhr, the maintenance and operation expense of 0,3 mills/kwhr, and the net power cost is 4,0 mills/kwhr. Substitution of sodium for LiF-BeF, as the intermediate coolant and replacement of the Loeffler boiler-superheater complex with a more con- ventional sodium-heated boiler would reduce the capital investment to about $125/kwhr. Development of alternative methods of processing (e.g., preci- pitation of rare earth fluorides) provides potential for reducing fuel cycle cost., An examination of the neutron economy indicates that the MSCR should be capable of e#olving_into a self-sustaining breeder reactor (BR *1,0) not dependent on outside sources of fissile isotopes. 8.7 Recommendations The comprehensive program of research and development for molten salt reactors in ppogress at ORNL is concerned at present with the construction and operation of the MSRE, This MSCR evaluation has disclosed certain additional areas of study, research, and development important to the realization of the nuclear and ‘economic potential of molten salt reactors. These arees are liéted_be;ofi,,together with specific examples in each area. 8,7.1 Title 1 Deszgn Study of MSCR Thls should be performed for two plant capacltzes (100 and 1000 Mwe) in order to detect unrecognlzed development problems and to verlfy the economlc predlctions made in thls report. The studyrpreferably should be conducted by an organlzat;on out31derthe Laboratory, but'With'Close liason and cooperation in special studies (e.g., nuclear design, reactor-integrated chemical'processiné)gf'A sffidyeof_lob'uwe installation is desirable to bridge the gap between the MSRE (10 Mwt) and a full-scale prototype. 174 8.7.2 Conceptual Design Studies of ‘Advanced Breeder Reactors Prior studies have established the'nuclear‘potential of a thermal molten salt-thorium breeder, However, this concept should be re-examined and developed in greater detail in the light of the technology accumulated in recent years. In additiofi, the potential of fast molten-salt breeders, including those breeding plutonium, or possibly both plutonium and 2330, should be evaluated. A number of concepts have been proposed, and it is not clear,'at present, which of these offers the greatest ultimate poten- tial coupled with the least difficulty-of development., 8.7.3 Fundamental Studies of Alternative Chemical Processes While the fluoride volatilify process is suitable fdr the recovery of isotopes of uranium from short-cooled fuel, its use alone does entail the discard of the carrier salt to rid the System fission products. This fact limits the processing rate and conversion ratio in thermal reactors using valuable isotopes of lithium and beryllium, although fast reactors using fluorides of sodium and potassium, etc., are not so limited. For the thermal reactors at least, an alternative process is needed in which sepa- ration of valuable components (thorium, uranium, lithium, beryllium) from fission product isotopes is effected by transfer between fluid phases, e.g., by extraction of molten salt fuel with a liquid metal, Driving forces for the transfer can be provided by the use of active metals and easily reduced fluorides or perhaps by the application of electric potentials, 8.7.4 Engineering'Laboratony'Study‘of'Precipitétion Processing This process, which has great potential for reducing the fuel cycle cost in the MSCR and which, by necessity, must be integrated with the operation of the reactor and placed within the reactor cell, could make ‘the introduction of MSCR power reactor plants independent of the availa- bility of central processing facilities for molten salts. This is an enormous advantage for the initial installatioms, particularly if these are widely scattered in remote locations. Also, the competitive position of the smaller installations (100 Mwe) would be improved. O ” # L] Y » 175 8.7.5 Pilot Plant Study of HF Dissolution Process This process, though not essential to the realization of the economic potential of the MSCR:would, if developed, make possible the evolution of the MSCR into a self-sustaining system, and would constitute a large Step in the development of two-region, two-fluid breeders. * 177 APPENDICES Introduction 9" In this portion of the report are collected reference materials, preliminary studies, and de- tailed discussions that support the assumptions : used or conclusions drawn in the main body of the report. Literature references in the Appendices do not refer to the Bibliography which follows, but to separate lists of references given at the end of each sub-appendix. ' et o s ek * 1 " 4« App 179 endix A MULTIGROUP CROSS SECTIONS FOR MSCR CALCULATIONS C. W. Nestor 34 Thermal Group Table A.1. Group Structure Group - Au u Energy (ev) 1 0.91629 0.916 4 X 108 = 107 2 0.69315 1.609 2 X 10% - 4 x 106 3 0.69315 2.302 1 -2 x 106 4 0.20400 3.506 3 x 10° — 10°% 5 1.09860 4605 1 x 10° — 3 x 10° 6 1.20400 5.808 3 x 10* -1 x 103 7 1.09860 6.907 1 x 104 - 3 x 10% 8 1.20400 8.111 3 x 10° -1 x 104 9 1.09860 9.210 1x 10 -3 x 10° 10 0.91629 10.126 400 - 10° 11 0.98083 11.107 150 - 400 12 0.40547 11.512 100 — 150 13 0.10536 11.617 90 - 100 14 0.11778 11.735 80 — 90 15 0.20764 11.942 65 — 80 16 0.26236 12.204 50 — 65 17 0.10536 12.309 45 — 50 18 0.19574 12.505 37 = 45 19 0.11441 12.619 33 — 37 20 0.09531 12.714 30 - 33 21 0.18232 12.896 25 — 30 22 0.22314 - 13.119 20 — 25 23 0.16252 ©13.282 17 - 20 24 0.23052 13.572 13.5 — 17 25 0.30010 13.813 10 - 13.5 26 . 0.28768 . 14.101 7.5 =10 | v 0431015 14.411 5.5 7.5 28 0.31845 14.729 b —.5.5 29 0.47000 15,199 2.5 2 4 30 0.57982 - 15.779 1.4 = 2.5 cil 0.55962 16,339 0.8 —» 1.4 32 0.28768 16.627 0.6 » 0.8 33 Ep. Thermal 0.31705 16.944 0.437 - 0.6 (2200°F) - 0.07940 180 Lithium-6 Table A.Z2. ggT Group AL AL LA NN NNy 2 2252935555 %5%%%%%%%%% X XAAXAAXAXAXAXAAXAXAXAKXAKXAXXXKXAXXX 3R AN NS RIREBRIEES AN IS ITRRRIRY OO0 O0OO0C00OC0COCO00COO0OO0O0OO0O0OO0O00O0O00O00DO0O00D0O000O0O0O0O00O00O0O0 ~ mlllll lllllllllllllllllllllnwwuu*l XXXXXX X.XXXXXXXXXXXXXXXXXXXXXX By 8RR AR RN R0 NgSRR8RAR 76514610606356791356Wl584u994l4827 OrdAN~HFANNAN O A A A NN N0 A o~ TTTET T | OO0 9855585 998223323338 XX XXX XX XAUAXAXAKXAXX X XXX VOOV ANINTYDOANANNN-HOD NN NN SOOI AN O A A NOOOMNANNVOWWOMWNWNEONOTARONNINWHOMONMNT N . - . . . [ L] . - . . . . 4 & o s & & & 5 s A QMNP D00 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 is 181 Lithium-"7 Table A.3. lo o Group NANAYVOOO0OO0000Q0 0000000000000 000O00Q0O0O0 T O A DN N NN N NN N NN N N NN N DN NN DN NN DD N N 0N *« o+ @ * @ *« 9 * & & 9 = = = a ® *® & & . . . = . @ . & * @ NN O NNANANNNNNNNANNNNNNNRNNNNANNNNNNNNN Y P T P Y ¥ YT T I YT T I N N N NIRRT YV OO0 00000CO00000O0OCOO0OO0O0000000O0O000O0O HA A A At A A A A A AAAA A~ KUYXAUAKAAKYAEXAIAAXAXKAARAXAALAXKHEXAXUYXXXNYXXXX XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX ~NNoaNNweR----- -~ ............ HNMN TN OO NN ONORNO A NM N ON0NO oM _ ll1Bllllll22222222223333% 182 Table A4, Be:yllium-9 Group EUT, Ua ch _ 30tr 1 0.4608 6.3425 X 10~7 0 5.2490 2 0.3240 9.4485 x 1077 0 6.6160 3 0.6813 1.3362 X 10~° 0 6.6919 4 0.8700 2.1688 x 108 0 1.0649 X 10 5 1.1913 3.8471 X 10°° 0 1.3588 x 10 6 1.2122 6.8585 x 10-° 0 1.6068 x 10 7 1.2122 1.2165 X 107? 0 1.6112 X 10 8 1.2122 2.1688 x 10-° - 0 1.6112 X 10 9 1.2122 2.8471 x 10~3 0 - 1.6112 X 10 10 1.2122 6.3425 X 1073 0 1.6112 X 10 11 1.2122 1.0204 X 10~4 0 1.6112 X 10 12 1.2122 1.4312 x 10~% 0 1.6112 X 10 13 1.2122 1.6236 X 10-% 0 1.6112 X 10 14 1.2122 1.7167 X 1074 0 1.6112 X 10 15 1.2122 1.8628 x 1074 0 1.6112 X 10 16 1.2122 2.0956 X 104 0 1.6112 x 10 17 1.2122 2.2960 X 10~% 0 1.6112 X 10 18 1.2122 2.4762 X 10™4 0 1.6112 X 10 19 1.2122 2.6752 X 104 0 1.6112 X 10 20 1.2122 2.8190 x 104 0 1.6112 X 10 21 1.2122 3.0224 X 10~ 0 1.6112 X 10 22 1.2122 3.3454 X 104 0 1.6112 X 10 23 1.2122 3.6832 X 10™% 0 1.6112 x 10 24 1.2122 4.0646 X 1074 0 1.6112 x 10 25 1.2122 4.6430 X 104 0 1.6112 X 10 26 1.2122 5.3775 X 10~ 0 1.6112 x 10 27 1.2122 6.2450 X 10™4 0 1.6112 X 10 28 1.2122 7.3085 X 10~% 0 1.6112 X 10 29 1.2122 8.1920 X 10~ 0 1.6112 X 10 30 1.2122 1.1601 X 10~3 0 1.6112 X 10 31 1.2122 1.5142 X 103 0 1.6112 x 10 32 1.2122 1.9013 x 1073 0 1.6112 X 10 33 1.24 2.15 X 10~3 0 1.611 X 10 34 1.24 5.048 x 10~3 0 1.611 X 10 L v 183 . Table A.5. Carbon-12 Group €9y T Vo, 3ty 1 0.303676 2.1565 X 10~7 0 3.7354 2 0.2686 3.213 0 5.4204 3 0.4108 b o 502, 0 0.0604 4 0.6241 7.380 0 9.2352 5 0.694568 1.308 x 1076 0 1.19114 X 10 6 0.74181 2.332 X 1076 0 1.28686 x 10 7 0.7584 4,187 X 107 0 1.35370 X 10 8 0.7584 7.375 % 10~ 0 1.35936 x 10 9 0.7584 1.308 x 10-3 0 1.35936 x 10 10 0.7584 2.157 0 1.35936 x 10 11 0.7584 3.470 0 1.35936 X 10 12 0.7584 4. 866 0 1.35936 X 10 13 0.7584 5.520 0 1.35936 X 10 14 0.7584 5.830 0 1.35936 X 10 15 0.7584 6.335 0 1.35936 X 10 16 0.7584 7.125 0 1.35936 x 10 17 0.7584 7.805 0 1.35936 x 10 18 0.7584 8.420 0 1.35936 x 10 19 0.7584 9.095 0 1.35936 X 10 20 0.7584 9.585 0 1.35936 X 10 21 0.7584 1.028 x 10™% 0 1.35936 X 10 22 0.7584 1.138 0 1.35936 X 10 23 0.7584 1.253 0 1.35936 X 10 24 0.7584 1.382 0 1.35936 X 10 25 0.7584 1.579 0 1.35936 x 10 26 0.7584 1.829 0 1.35936 x 10 27 0.7584 2124 0 1.35936 X 10 28 0.7584 2.485 0 1.35936 X 10 29 0.7584 3.030 - 0 1.35936 X 10 30 0.7584 3.944 0 1.35936 x 10 31 - 0.7584 " L. 5.245 0 1.35936 x 10 32 0.7584 C 6.Ab5 0 1.35936 X 10 33 7.585 x 10"1 1.85 x 1073 0 ©1.359 X 10 34 - 7.585 x 1071 5.048 x 1073 0 1.359 X 10 184 Fluorine-19 Table A.6. . EO_T lo Group 0 X 10 x 10 X 10 X 10 X 10 x 10 oo o o . (oR® 99939382333 83832883331 XX XXX XXXXXXXXXXXXX X 0005005u111 NN NN DD DD DD D00 DD NN 000N NN O M SNy 6 ~ o~ O0O000O0O0COO0OLOO0O0OO0OO0OO0O0O0COO00O0 UL . . . . . I L A - T . I P P R P R R e L L L X 10 X1 2995 X XXX 00000000000000000000000000000000,0nU ~- ,333 © o ©o©o of — = . | A9 XOOXXXO0O000000000OO0O00OO0O0O0O0O0O0O00OOX Q QOQ | , o o SO0S 0 10 o ~ o — ~ A AAAAAdAededaMAM"AddAdAdAdAAAAdSAASA~AAA« 55558655656656b5556556060088008800066099 AAAdAAAddAdAAAdAdAAAAAAAAAAAAAAAAAAAAAAS XXXXXAXXXXXXXXXXXXXXXXXXXXXXXXXXXX NONOONINOO0ODOODO00O00O0O00O00000DO0O0OQO0O0ODO 0O 6045753777775555555555555555555555 lllll . . . . - . - e . . . N N - r T Y ENEN MY N M Y M M MM @M Ea e Y~ 0O ~ - -N " 185 Table A.7. INOR-8 goT ca Group £ tr 1 5,98 x 1071 1.868 x 1071 0 10.347 2 5.89 x 1071 1.904 x 10-1 0 10.693 3 6.98 x 1077 1.936 x 107t 0 10.521 4 1.010 1.972 x 1071 0 10.760 5 2.375 2.006 x 10 0 13.270 6 3.158 2.031 x 10~% 0 14.48 7 3.165 2.057 x 107} 0 20.793 8 3.175 2.096 x 1071 0 42.011 9 3.010 : 2.264 x 1071 0 48.153 10 2.939 2.544 x 1071 0 54,.372 11 2.930 3.531 x 10-% 0 59.116 12 2.941 5.778 x 10"1 0 57 .442 13 2.967 1.159 0 56.977 14 2.967 1.159 0 57.488 15 2.967 1.159 0 57.58 16 2.967 | 1.159 0 5762 17 2.967 1.160 0 58.23 18 2.967 1.161 0 57.30 19 2.977 9.800 x 10~? 0 56.00 20 3.018 2.009 x 10°? 0 57.070 21 3.018 - 2.016 x 101 0 56.140 22 3.018 2.027 x 10-1 0 57.349 23 3.018 2.043 x 1071 0 58.698 24 2.995 2.092 x 10°% 0 57.628 25 2.978 2.291 X 10°1 0 58.698 26 3.077 2.652 x 1071 0 58.837 27 3.077 . 3.038 x 101 0 58.930 - 28 . 3.156 - 3.554 x 10~% 0 59.116 29 3.270 . 4.215 x 1071 0 58.279 30 3.270 . 5.099 x 10-1 0 58.279 31 3.141 .. 6,740 x 1077 0 57.349 32 '3.152 . 8.809 x 1071 0 57.74 33 3.285 '1.258 0 58.695 34 0 58.693 186 Xenon-135 Table A.8. lo Eon Group e 0000000000000 OOO0O0OOOO0OO0OO0OO0OO0O0O00OO0 000000O00.0000000000000000000000000 198 x 10% 198 x 10° 198 x 107 050 x 102 282 x 10° 282 x 107 222 x 107 254 x 10° 254 % 102 254 x 10° 254 x 107 7.035 X 10% 7.43 X 1072 3.908 x 1071 7.242 x 10°1 '9.843 x 1071 9.546 x 101 3.960 x 1071 1.515 x 10 3.045 x 10 5.133 x 10 g8.580 x 10 1.047 X 107 1.870 x 1072 6.781 x 10?2 2.439 x 103 2.439 x 103 g8.561 x 103 1.720 x 10* 1.720 x 10% 2.100 x 10% 5.697 x 10% 4.20 x 10% 1.60 x 10° 2 2 2 3 3 3 4 8 8 8 8 D000 O000000000000000000000000000OOO HNOAOFT N0 N~ OO NN OO HANM llRlll111122222222223333% 187 Group gcT Ua vcf 3°tr 1 0 3.000 x 10-1 0 0 2 0 3.000 x 10-1 0 0 3 0 3.262 x 1071 0 0 4 0 3.600 x 10°1 0 0 5 0 1.027 0 0 6 0 1.570 0 0 7 0 1.570 0 0 8 0 2.240 0 0 9 0 1.641 X 10 0 0 10 0 4.857 x 10 0 0 11 0 4857 X 10 0 0 12 0 8.829 x 10 0 0 13 0 1.911 x 102 0 0 14 0 1.911 x 102 0 0 15 0 1.911 x 102 0 0 16 0 1.756 X 102 0 0 17 0 1.714 x 102 0 0 18 0 1.714 x 102 0 0 19 0 1.586 x 10° 0 0 20 0 1.037 x 102 0 0 21 0 1.037 x 107 0 0 22 0 1.037 x 102 0 0 23 0 1.037 x 107 0 0 24 0 1.463 x 102 0 0 25 0 1.887 x 102 0 0 26 0 5.991 x 107 0 0 27 0 5,991 x 107 0 0 28 0 5.991 X 107 0 0 29 0 1.000 x 1072 0 0 30 0 1.000 x 10% 0 0 31 0 1.391 x 10° 0 0 32 0 4. 774 % 103 0 0 33 0 7.95 X 102 0 0 34 o 4.20 x 10% 0 0 Table A.10. Samarium-151 188 Group g, < Q g‘O’T a tr 0—26 0 0 0 0 27 0 5.50 x 102 0 0 28 0 0 0 0 29 0 1.30 x 1072 0 0 30 0 7.20 x 102" 0 0 31 0 2.28 x 103 0 0 32 0 7.25 x 107 0 0 33 0 9.80 x 10° 0 0 34 0 4.91 x 10° 0 0 Table A.1l. Thorium-232 (Infinite Dilution)* Group Eon Oq fifif BEer 1 6.100 x 1072 0.2408 5.642 X 1071 1.883 x 10 2 5.669 X 1072 0.1684 3.190 x 1071t 1.998 x 10 3 5.841 x 1072 0.1450 1.306 x 1071 1.966 X 10 4 8.160 x 1072 0.1622 0 2.366 x 10 5 1.039 x 10-1 0.2708 0 3.132 x 10 6 1.119 x 10™% 0.4627 0 3.718 x 10 7 1.130 x 1071 0.5680 0 3.739 x 10 8 1.156 x 1071 0.7958 0 3.739 x 10 9 1.168 x 107t 1.029 0 3.739 x 10 10 1.180 x 1071 1.168 0 3.739 x 10 11 1.339 x 1071 7.234 0 4.658 x 10 12 1.824 x 1071 20.802 0 6.349 x 10 13 1.202 x 1071 1.481 0 3.739 x 10 14 1.204 x 1071 1.502 0 3.739 x 10 15 2.301 x 107t 82.755 0 8.040 x 10 16 1.728 x 1072 16.568 0 6.014 x 10 17 1.213 x 1071 1.615 0 3.739 x 10 18 1.217 x 107% 1.646 0 3.739 x 10 19 1.219 x 1071 1.678 0 3.739 x 10 20 1.221 x 107} 1.701 0 3.739 x 10 21 1.224 x 1071 1.731 0 3.739 x 10 22 5.117 x 107t 198.0 0 1.781 x 10 23 1.232 x 1071 1.819 0 3.739 x 10 24 1.236 x 1071 1.864 0 3.739 x 10 25 1.242 x 1071 1.927 0 3,739 x 10 *¥Groups 11, 12, 15, 16, 22 computed in Appendix B. O Ufi 189 Table A.11 (continued) £0q Group QOO0 0OO0O0OO0O - llllllllm COQOTCO0OOO0O0O0O0O0O000O0O0O0O00000O0000 0O xxxxxxxxx HA A A A A A A A A A A A A AAAAAA A A Ao A OO O 8 XAXAKXKXAEAKAXAAAAKXAXAXYXAXAXAXXAYXXYXX XX MmO Mg MmOy - -+ ‘ - lo WO AAdt AT At A A A A A~~~ =~~~ e+ s s s s e s (ea) RN OO AN YO MMM~ ~F TN S SN N NN NN NN OO OO OO OO ON OO o0 o % _w COO0OCO0O0CO0O0O0O0O00OO0O0D00O00O0O00O0O00O0O \ _ OO0 O0O00O0O00 m o s o o 42 QOWONDO W : Q NN OANNOO AN ONHMANOODOO a3 a NMNMNNO~NTNONNOOVNNONAONNOADIND VOO 2 lo QOO0 ANVWOWNETONNHINOONOOAH O OO IN OO o _ OCOO0O0O0ONOANMNYVOERE-OHOOAHNMODO M ¢ ® & *» ¢ &+ & « & 8 5 B 8 s & 2 @ @ 8 e e « & = @ O WOVO O~ ~ N H COO0O0O0CO0O0O00O0O00CO0O0O0O0O At A~ HNN OOV OVONO N~ D - , S ] OO AN~ Ot~ _ _ . o AN AN AN AN TN NN OO R ) _ < NN NN NN NN NN NN N e A A A A A o bbbbobbbbobbbbbobbbbbbbbbbhls dddflqqqfld fl A A A A A A A A A A A A A A AT A A A A A A Ao OCO0O00OO0O0O0O0O ® , _ > , XXX XXX XXX VOV OO ANNANOOAOOO0ONO A NT OO F O PO M b O st lo Eo Group OO OO0 O0O000OO0000000COO0CO0O00 ™ W WY Wy 0 W0 W0y 0 W00 DY D 00D W1y 0y vy iy MONNANNNNNNNNNNNNNNNNNNNNNNNN NN N OO HEOWV N NN AN ININN NN OO O DD 0D 00N ~ MOVINMANNNANNANANNNNNNNAAAAANOAOO MANNNNNNNNNNNNNNNNNNONNN NN NN NN~ 2 D — — X % % OCNHNHHOAOOOOOOOOOROONOONOOD . BM9673993333333333333333300000 N~ AN N Hred A ANNQANNOOAO OO OOOOEM OO - cooocooBooBoodBS N N (N N N ™M AAAAAAAAAAAAAAA9999509979 xxxxxxxxxxxxxxxxxxxxxxxx OO0O0O0OO0OHWINNINNANMNMRND D0 O A0 HOO~NOANMNMOOTN~ANNTOONHOYONEOMO A NHNAHAMOAAOAQONDOOUNTITAR O FADORO VNNV AMNNNTOO A A0 N AN NE AN A A o000 ddd oD o llllllllllllllllllllmmm XXXXXXXXXXXXXXXXXXXXXXX OO0 OONWWHHNHOANNONOAODNNWORNOO 41561669013338517247324E5m0869 RO H OV OO HVOOMNVOVOANDONAHOOSMAN HreHd N NN NNFAN AT NN DO O A NANN NS00 N NN AN AN A A A A A A A A A A A A - - A A e bobbboOLbbObbLLLOLLOOOOOLLOLOLOLLOLOL HAAAAAAAAAAAAAAdAdAAdAd dAAAAAAAA HUYXXHEAUMYXKAXAEARAXXAXAXAXAHYAKAXAXXAYYXXXXXX D000 O0OMNAH YO HNNNANNHAHOETFTONRALTOOOOO 0152467632052775592&3581&5m515 THOAOAANAMMOLTALTNONANNANTINTOANONOOND ™ VO AANANMETNIN NN OO YOO HNALTNOTONO A NN OO AN OO llulllllll2222%2%2223 & 191 Table A.13 (continued) G.roup EGT Ua ' Vdf 301;1‘ ' M v 31 24740 21.79 x 10° 3.990 x 10° 5.40 x 102 2.23 2.50 32 1.430 1.405 x 10° 3.217 x 102 3.739 x 10 2.29 2.50 33 1.545 1.27 X 10?2 2.9 x 10% 5.40 x 102 2.29 2.50 34 1.545 3.40 x 10 6.54 x 102 8.47 X 10° 2.29 2.50 Table A.1l4. Uranium-234 Group g0y % Vo, 30 1 7.298 x 10~% 1.556 3.875 1.754 X 10 2 6.865 X 10~% 1.529 3.800 2.019 X 10 3 6.718 x 1072 1.353 3.350 1.990 x 10 le 8.020 x 1072 0.673 1.630 2.347 x 10 5 1.007 x 1071 0.083 0.121 3.176 X 10 6 1.122 0.067 ' 0 3.709 X 10 7 1.188 0.118 0 4.011 X 10 8 1.167 2.104 0 4,104 X 10 9 1.234 - 0.373 0 4.187 X 10 10 1.258 0.615 0 4.187 X 10 11 1.044 9.054 0 3.290 X 10 12 1.068 2.077 x 10 0 3.290 x 10 13 1.075 | 5.957 X 10 0 3.290 x 10 14 1.083 5.486 0 3.290 x 10 15 1.099 1.833 x 10 0 3.290 X 10 16 1.122 S 2.109 x 10 0 3.290 x 10 17 0 1.132 7,901 x 10 0 3.290 x 10 18 1.152 24402 - 0 '3.290 x 10 19 1.165 - 2.595 0 3.290 x 10 20 1.176 2.690 x 102 0 3.290 x 10 21 1.199 . 2.932 0 3.290 x 10 22 1.229 . 3.245 0 .3.290 x 10 23 - 1.254 - . 3.573 0 ~3.290 x 10 24 1.293 3,943 0 3.290 x 10 25 1.350 T 4.504 0 3.290 X 10 26 1.244 1 5.216 0 2.692 x 10 27 1.324 . 6.509 x 102 0 2.692 x 10 28 1.420 - 1.158 x 10° 0 - 2.692 X 10 29 1.593 8.645 : 0 2.692 X 10 30 1.871 1.125 X 10 0 12.692 X 10 31 2.228 1.496 x 10 0 2.692 x 10 32 2.4.54 1.844 x 10 0 2.692 x 10 33 2.749 x 1071 2.144 x 10 0 9.104 X 10 34 2.749 x 1071 5.54 X 10 0 2.20 x 102 192 Uranium-235 Table A.15. lo €0, Group XKAKAAXAKXAAKXAXAXXXXKXX N OO NOINT T PP P N PP I T I PPN P Y Y 2/»....._)._.299777777777777777777777777726 AAAANRAOAMOOOMMOMOMAMO®EM@ANMOA@ @O @ o 10 10 - 10 0000000t obbBoooo A SAa A AddAddASSa9a3as XX KEUYXXRXXXXXXXXRXXX XX XXX XX X 508D A -0y iy~ (N ) ~t Oh O NO~SE~EFFN~T OO NN NO OOV EHVOONANNOO 511E83ll 467090140 92043364504492 32_3335811._2«454761992181281272741116 o N N NN o O o 2293223333%23%%83%%23332329% XXX KXAXXXAXAXXAXXXUXXUYAXXXXAXAXXAXXXX ) QOO0 O I~ WO -0 00 0Oy ONWE-rded 000 N H MO D~ IN 8888881 3967778 455990412M09232120 o 4 YT YT YT YT YT Y YT T T YT T T YN T YT | . 1 o O Q o 98858555555855855855525855555558558545 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 2113043368 62.4 09230199775 61770120.43 00N MSR49408832924@RR 665711122232333654325344212223671l HNMENON0oOg NN 193 1 Uranium-236 Table A.1l6. Utr €0 Group 1.754 x 10 2.019 x 10 2.965 2.242 1.592 1.196 9.344 x 1071 1.990 x 10 8.993 x 10°¢ O00CO000O0O0CO0O00O0O0ODOOOO0O Q0O OO lmlmlllmllllmlmlllllmmlllmmllmm xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx O ANDDO00000CO0O000000O0O0NANNNNN N N O OO0 ARAERIORRON OGN IN MAO A AANNNNNNNANNNNNNANDOOOO 0000 2334444333333333333333222222233 o ) — X 5000000000OOOOOOOOOOOOOOOOOOOOO N ™M 3 — - o~ N A AA - - TrTTTITY T cbobboooo b5LLLEELS O0000O00O0 A A HAMA A A3 4~ 1lllllllxxxxxxxxx X X X X XXX X X XX XX X XXX - 77225228504268091 WROO NG ~F 3 VOANNNONNEENNNOHAHMNQSM O PWOVVNNONFAOACANNOAITNNONDNONAO M A 3248124711811977172223333681112 2 194 Neptunium-237 Table A.17. toy tr Group OCOO0OO0OO0OO0COO0OOO0O0O0 Table A.18. Uranium-238 lo E0p Group > oo oCcoC O 00 0 9999999292223 223338353322° HHYAXXARXRYXKEARXAYXXAXXXXXXXXXXXX NNV AR O OO ANMNO A A ANANANNNNNNNNNNANNNANN — .O ~ X 254000000000.0000000000000 OO ~ ~ O - A A A A A 0565006609 o 9o S 9 A A A A A ~ ~ ~ r~{ = XXX XX XXX X X X X X ~r VO NOOONOOD AT ONOY O OO0~V 7%43932713564192228227224 YY Yy TTY T TYd T Ye Tt TN 00000 CO OO 0O © OO 9399922232338 32888528583R33R5 HHUYAKXAHKXARAXXXAKXAXAKAXXAXXAXX AKX XX oo HdNANDOAMMOMAaMMAaMMOaMMeamMm MNMONOFNANNNNOOOOOOOOOOOOOOO OO BN A AAdAAAdAAAAASAA S AAAA NNt O~ 00O M~ OO ANMSE N 11R1111111222222 . 195 Table A.18 (continued) E0m Group OO000QOOO0OO0 AAAAAAAAA XXX XXXXXX OO0 OO0 I O O O OO O YWY Nl i o o i o e ot e OCO0O00O0OOCO0O0 - Y O = — Y X WO AN O NN OO M AN A A A Ao N NN NN U L1 OO0 O0O A A A A e X X X X X X X XX O O WO O\O OO ..... ; M _ 196 Appendix B EFFECTIVE THORIUM RESONANCE INTEGRAIS - J. W. Miller#* Introduction The "effective"” resonance integral to lethargy um of a particular isotope in a mixture of isotopes is defined by the equation L . I (eff) =f - ("ao')eff du - O s The value of (oao)eff is a complex function of a number of variables. A most important item is the scattering power of the medium, measured by op, macroscopic scattering cross section per thorium atom, o, = fi'Nko;/Nth . The greater the scattering power, the greater the probability tha£ a neutron will be slowed through a resonance without absorption. The thermal motion of absorber nuclei, on the other hand, increases the probability of absorption due to Doppler broadening of the resonances. This effect is small at 20°C, the standard temperature for reporting measured values, but increases at higher temperatures. Dresner has developed a method for com- puting the temperature effect for resolved resonances. > Group averaged cross sections for thorium are related to the effectiVe resonance integral and were calculated by means of the equation ui +-Aui i T M (000 )epr 0¥ where i is the group number. ¥Adapted from ORNL-CF 61-1-26. wi 197 Analzsis By assuming a Maxwell-Boltzmann speed distribution for the moderating medium and the Breit-Wigner formula for a single resonance, Dresner? has arrived at the following equation for the resonance integral, Ieff,for a single absorption resonance: T g, I'y 28 Ieff_— — J{¢&, k): 2 E T o where s . 4mx’ I'm o r ’ o B='f'2: o o T §=Z, > l = = 2 — . o Other terms are defined in the nomenclature. The function J(£,k) is tabulated in ref. 1. The tabulated results have been plotted as the family of cuives in Fig. B.l. The reduced mass of the neutron p is equal to neutron mass (~1.0) multiplied by M/(M+l), where M is the mass of the absorbing atoms. Since the mass of thorium is 232, (o was taken as unity for these calculations. Table B 1 llsts the resonance parameters: for—each ‘of -the 13 resolved thorium resonances (2). These parameters were used in the equatlon for I of f for computlng the resonance 1ntegral for each resonance. (Slnce these calculatlons were performed additional thorium resonances have been re- solved, and 1mproved"parameter values for_prev1ously resolved resonances. have been obtained. The effect of these new values has not been evaluatéd) The total resonance integral for each energy group is then obtained by ad- ding together the separate integrals for each resonance. Table B.l also 3 (&, A 198 ORNL-LR-DWG. 54604 0.4 0.2 0.3 0.4 - 0.5 0.6 f . Fig. B.1. J(&,k) Versus & for k = 6.16. 0.7 C X 199 Teble B.l. Resonance Parameteré of Thorium-232 Infinite GNU | Dilution Group No. ou Eo(ev) Fn(ev) Py(ev) Resonance Integral 235 0.0017 0.034 234 0.027 0.034 _ 212 0.0012 0.034 1 0.98083 5;1 0.016 0.034 7-10 195 0.022 0.034 172 0.067 0.034 130 0.009 0.034 12 0.40547 122 0.023 0.034 8.4 114 0.013 0.034 15 0.20763 69.7 0.039 0.043 17.2 16 0.26236 59.6 0.0046 0.021 AN 23.6 0.Q04 0.039 22 0.22314 51,9 0.0022 0.034 42 Unresolved Resonances — Group 1 through Group 10 5.3 Nonresonance Contribution — Groups 13, 14, 17-21, 10.3 and 2333 Total — Cut-off at 0.0795 ev 96.9 lists the lethargy widfh'for'eachrgroup. The group-averaged cross section is simply the group total résbnance integral divided by the group lethargy width. | | o | - The infinite-dilution resonance integral (I_) for a particular - resonance may be obtained from 200 Sample Calculation Example: Compute the effective resonance integral (Ieff) for the thorium resonance at 23.6 ev with a carbon-to-thorium ratio of 200 and an absorber temperature of 649°C. ‘The ndhresonance scattering cross section, USC, of carbon is 4.8 barns, and the thorium nonresonanée scatfiering cross . Th . section, 0, » 18 12.5 barns. 972.5 barns . o ot I 2 2 2.86 x 10-° - 2 1 VE 0.877 X 107°0 cm? . ' 411?C2I',— W Q " 10,250 barns . i o 4. B=_B 9] o 0.0949 . 1/2 5. A=(M) M 0.1797 . o ur I >+ 0.2393 . 201 _ _ In g +5 1n 10 | 7 k_ 1n 2 = 13-2 8. J(&,k) = J(0.2393, 13.2) 7.6 {from Fig. B.1) . fl T 0, ' 2B 9 Ieff =-'—"-'Z"—'J(§Jk) 2 E: T o = 12.2 barns . The infinite-dilution resonance integral (Im) obtained is 26.6 barns for the case considered. Results The MERC-1 program used in the MSCR study is a 34-group diffusion- theory code. The group structure is such that the resolved resonances of thorium fall into five groups: 11, 12, 15, 16, and 22. The group-averaged absorption cross section for_eagh of these five groups as a function of cp at 649°C is plotted in Fig. B-2. The total resolved resonance integral is plotted ih Fig. B-3 for thiee,reactor temperatures. Symbols P7J= Radiative capture width (ev) I'n = Neutron width (ev) | T = Total width (ev) | x = Wavelength gfthe neutrqn (cm) E = Resonance energy (ev) oq (barns) 202 ORNL-LR-DWG. 54606 GROUP 1 GROUP 1 GROUP 1 GROUP 11 4 6 8 102 2 4 6 8‘03 2 4 op (barns) Fig. B.2. Group Averaged Absorption Cross Section Versus Op 649°C. O O o B A e — RESOLVED RESONANCE INTEGRAL (barns) -l o 0 o ORNL-LR-DWG, 54608 T T 1111 l | — CURVE TEMPERATURE, C A V" A 925 /// B 649 .~ c 280 /pfi 7~ 7/ g JVA DO S Z //¢V ” _37/]1 __‘C/ | _ 4 6 8 102 2 4 6 8 103 2 4 6 ‘o (barns) Fig. B.3. Total Resolved Resonance Integral Versus Op for Moderator Temperatures of 280°C, 649°C, and 925°C. 3 * 8 2w ] 204 Nonresonance scattering cross section (macroscopic scattering cross section per thorium atom density) Relative energy in the center-of-mass-system Mass of absorbing nucleus (amu) Absorber temperature in energy units (ev) Reduced mass of neutrons (amu) References Lawrence Dresner, Tables for Computing Effective Resonance Integrals Including Doppler Broadening of Nuclear Resonances, USAEC Report ORNL-CF 55-9-74, Oak Ridge National Leboratory, September 19, 1955. D. J. Hughes and R. B. Schwartz, Neutron Cross Sections, USAEC Report BNL-325, Brookhaven National Laboratory, July 1, 1958. o C ;gh‘-;__‘ 205 Appendix C ' ENERGY DEPENDENCE OF ETA OF 233y# C. W. Nestor Summarx A paramefier of great interest in nuclear calculations of a thorium reactor is 1, the number of neutrons produced per neutron absorbed. Ex- perimental information on the energy dependence of 7 in the range of O to 10 ev as measured at the MIR was used to calculate group averaged fis- sion cross sections, using absorption cross sections calculated from the recent total cross section data'’? and the scattering cross section as calculated by Vogt.! The values used in preparation of the cross sec- tions was normalized to a 2200 m/sec value of 2.29.3 In the energy range of O to 0.8 ev, n was assumed to be constant at 2.29. 1In the range of 0.8 ev to 10 ev, as mentioned, group averaged values of'VEf =‘fi3a were calculated by numerical evaluation of the inte- fn(E) o (B) & greals 'T-‘. - g = | (E) a Au where Au denotes the-lethargy width of the group. In the range of 10 ev to 30 ev n was estlmated to be 2. 17, the data i of Gaerttner and Yeater® indicate an average 7 in this range of about 0.95 times the 2200 m/sec value. *Adapted from ORNL-CF 61-6-87 (Rev). 206 From 30 ev to 30 kev, 7 was assumed to be 2.25; this is the value - reported by Spivak® et al., at 30 kev. From 20 kev to 900 kev, measurements of 1 are available;? fission cross sections are reported in BNL-325 for the range 30 kev to 10 Mev. The total cross section in this range was taken to be equal to that of 23"E’U, as suggested by J. A. Harvey.? The value of Vv was assumed to be linear in energy with a 2200 m/sec value of 2.50 (ref. 3) and a slope of 0.127 per Mev.® A plot of the experimental n and the group averaged values from 0.01 ev to 1 kev is shown in Fig. C.1. Group values are also listed in Table A.1l of Appendix A. O O ) "ApN3g ¥OSW Ut pesn (3).,U Jo sentes dnoan Tvp *9ig (A9) 3 0001 001 Ol o't I'o 100 _ _ ' T €l . ___ 'l o t - L n._ I il ) _ -19°| 1 “_ - “ - — . h.— | | ! ! \ I _ | =8’ | | | . D~ O . d I . 61 ; ) a A " "..:. =102 ni ' s \ _ - 1o 'y Vi & _ \ <4 12 Ny _"_ i ! \ - 1 i “ _p b P Y -2 v Iy . - I ' \.. i I | v .~ Y __. i _ ' \ -1 i\t | U 1 . - G'e Y §489% OMO-47-"INYHO (senpA paboaiaay-dnoig sy) seu)q plos ¢ (3)U 4o uoyoriop ey; 8jowys3 seul paysonQq) (B 208 References J. A. Harvey, ORNL, personal communication to C. W. Nestor, ORNL, March 1960. ' ' - ' M. S. Moore, MIR Nuclear Physics Group, personal communication to C. W. Nestor, ORNL, March 1960. J. E. Evans, reported at the Argonne National Laboratory Conference on the Physics of Breeding, October 1959. E. R. Gaerttner and M. L. Yeater, Reports to the AEC Nuclear Cross Section Advisory Group, USAEC Report Wash-194, USAEC, February 1958. P. E. Spivak et al., Measurement of Eta for 233U, 233U, and ?3°pu with Epithermal Neutrons, J. Nucl. Energy, 4:70 (January 1957). G. N. Smirenkin et al., J. Nucl. Energy, 9:155 (1955). 209 Appendix D THE MERC-1 EQUILIBRIUM REACTOR CODE T. W. Kerlin Introduction The MERC-1 code automatically calculates the composition of a fluid- fuel reactor so that equilibrium and criticality conditions are simul- taneously satisfied. MERC-1 uses the MODRIC' multigroup-diffusion-theory code and the ERC-10 equilibrium reactor code as chain links. These codes were previously used separately at ORNL in iterative calculations to de- termine equilibrium reactor compositions. MODRIC was developed by the Central Data Processing Group at the Oak Ridge Gaseous Diffusion Plant as a replacement for GNU.? The ERC-10 code was prepared as a later version of ERC-5.2 The MERC-1 code automatically transmits necessary data from one chain link to the other until sufficient iterations have been performed to cause criticality and equilibrium requirements to be satisfied simul- taneously. The output consists of equilibrium concentrations, a neutron balance, and fuel-cycle costs. The system which is considered in the MERC-1 analysis is shown in Fig. D.1. A complete specification of data required for controlling the flow and losses in each stream is included in the MERC-1 input. Note in Fig. D.1 that material is removed from the reactor system by losses, waste, and sale as well as by nuclear transformation (deéay and neutron absorp- ‘tion), and that fresh material is fed to the system. Therefore, the cal- culated e@uilibrium cohcentrgtionsrare in equilibrium with respect to the feed and discharge rates as well as with respect to nuclear transformation rates. Theory MODRIC is a typical neutron-diffusion-theory code. It allows 50 neutron energy groups with downscattering from a group to any of the 210 ORNL-LR-DWG 76824 Stream | ':_ Stream | | B Losses ' | © | Processing |—P== Waste *—%— | - & |_Plant ——P Sales Feed ., , | * Recycle 2-I - e Reactor* | .‘ : * ¢ P Recycle |— 2 Feed & | *2 O §‘ Stream2 |Pm= Loss | @ Stream 2 | Processing | g waste Plant ' Sales ®The reactor may have as many as two active regions, and each region may contain either or both of the fluid sfreams Fig. D.1l. The Reactor System. 211 following ten'greups. It will perform concentration searches on specified elements.' The output'cOnsists of criticality search converged concentra- tions, group macroscopic cross sections,rnormalized'nuclear events (ab- sorptions, fission,leakage, etc.) by region and group, absorptions and . fissions by material and region, group flux distributions, and fission density distributions. ‘ ERC-10 requires extensive input. Rewriting of large quantities of input is avoided by using basic input decks;which include information ap- plicable to a number of cases: To specify a new case, it is necessary only to spe01fy changes in the basic 1nput deck. For instance, one might prepare a basic input deck for a partlcular reactor w1th a given power level. A set of cases with dlfferent power levels would need only the basic input deck and the new power level as input. Basically, ERC solves two equatlons. They are: ~N,, (t,.+d,, +q,.—-r,.)=0, (1) or if the material must be fed to maintain criticality snm - s8-1 a . (N. » C N. . C' - ij 13 ij Tijk ‘ 1jk Z -1 f | | 13k 131‘ CR 2 Equation (2)'is Just thedcohsefvation requirement; saying that enough fis- '"_sile material must be added (or: removed) In 1teratlon. s to overcome the flneutron productlon def1c1ency (or excess) in 1teration (s-l) - These are ';inner 1terat10ns in ERC. The terms are deflned as. vy = volume of stream J, cm?," Nij = atoms of materlal i per “barn cm of stream j, t = time, sec, 212 Q.. = feed rate of material i into stream j, atams/Sec, R.. = rate of growth of material i in stream J due to recycle from other streams, atoms/sec, Fij = rate of growth of fission fregment i in stream i, atoms/sec, Ti' = rate of growth of material i in stream j due to neutron ab- 4 sorptions in other materials, atoms/sec, Di' = rate of growth of material i in stream j due to radloactlve 9 decay of other materials, atoms/sec, - | tij = rate coefficient for loss of material i in stream j because of neutron capture, atoms per sec per atom/barn-cm, di' = rate coefficient for loss of material i in stream j because J of radioactive decay, atoms per sec per atom/barn cm, qij = rate coefficient for loss of material i in stream j because of processing removal, atoms per sec per atom/barn cm, = rate coefficient for growth of material i in stream j because ij X J of recycle from stream j, atoms per sec per atom/barn cm, vi = neutrons produced per fission in material i, f - - - - . » - * - Cijk = reaction rate coefficient, number of fissions in material i per atom/barn cm in stream j in region k per fission neutron o born in reactor, i3k = reaction rate coefficient, number of absorptions in material i per atom/barn em in stream j in region k per fission neutron born in reactor. Superscripts: i = material, J = strean, = region. The use of stream and region indexes allows reactors wfith two streams in the same region to be analyzed. ~ The equilibrium concentration calculations in ERC use ieactiOn'rate coeff1c1ents (C, jk) obtained from an earlier MODRIC calculatlon. However, the 1n1t1a1 concentratlons used in the MODRIC calculation will not, in -~ general, agree with the equilibrium concentrations computed by ERC. This new set of concentrations will alter the neutron spectrum and flux level, thereby changing the reaction rate coefficients. 'Thefefore, it is neces- sary to repeat the MODRIC critieality caiculation with the latest value for the estimated concentrations to-get new reaction rate coefficients. - by dividing by N, the stream concentration 213 This process is repeated until the MODRIC and ERC concentrations are equal. The flow of information in the code is shown in Fig. D.Z2. The reaction rate coefficients (Cijk) used in ERC are spectrum-averaged cross sections which are available directly from MODRIC. The MODRIC cal- culation gives A_. and VF ‘the absorptions and neutron productions in ik ik? _ material i in region k, normalized to 1.0 total neutron produced. The distribution of nuclear events between two streams in a region is accom- plished by 1ntroduc1ng the stream volume fractions, fjk’ in this manner: A _ _(atoms of 1 in stream j in region k ijk = ik atoms of i in region k F _ (atoms of 1 in stream j in region k Vifise © Vifik atoms of i in region k The multiplying factor in each term is (atoms of i in stream j in region k - 13 fgk atoms of i in region k Z N. ' ij Jk where the units on these factors are _atoms of i in stream j ij = Dbarn ecm of stream j ? - em® of stream j in region k £ = ' . d - cm? of region k The material, stream, and region”dependent absOrption and production terms are automatically transferred from the MODRIC link to the ERC llnk of the - MERC-1 calculation. ERC obtains the reactlon rate coefficients (1nten51ve' . quantitiee) from the absorption and productlon terms (exten51ve quantities) 214 ORNL -LR-DWG 76825 First Guess at Concentrafions MODRIC__ __ r—-————"—"—"\|1""—"——-—-= =1 | | I | | | | - Calculate Keff | | & IsKeff= 1.0 | 1 NoA Yes | | [Change Concentration| I L - GwEe ouED eums GEmn sues cvus e s caE GEES SR GEED GEED GEREED GEED GWSD S —l ERC 4 | | l | | | Do these two.concentrations | agree | - b | No Yes | Do MODRIC and ERC agree on all concentrations No Yes Is moximum number of iterations exceeded No Output Output Fig. D.2. The MERC Flow Diagram. The absolute reaction rate coefficient C. .. 1s obtained in the ERC cal- ik 7 ~ culation using the total neutron production rate as determined by the re- actor power Ef‘jk = 131«: X 3.1 X 10® P ¥ X 10724, where 3.1 X 10*® = number of fissions per sec per megawatt, P = power level in megawates, V= average number of neutrons produced per fission ZN..C?. v ] i 3k ij "ijk i E f N,. C,. i3k i "ijk A similar argument applies for the fission reaction rate. The restriction to fluid-fuel reactors occurs because thé ERC calcu- lation requires that the reactor discharge composition be equal to the mean reactor composition Thls restrlction is satisfied in fluid-fuel re- actors but not in solld-fuel reactors where the dlscharge has experlenced a much greater exposure than the mean. In ERC-10, effectlve, one—group cross sectlons of 1nd1v1dual fission products were calculated by reference to & standard absorber exposed to the MSCR spectrum of neutrons as generated by the multigroup program Modric. ‘The thermal cross section endfithé,reSOnancefintegral can be used in a two- group'model to calculate,relatiVe'one-group Cross sections. - Thus, set where the bar denotes effective, one-group values, "e" denotes "eipthermal," and "t" denotes "thermal." Rearranging and taking a ratio of the cross 216 section of the ith material to that of a standard material, denbted by the " subscript "s, al ———— Q ct - ‘&I‘& ct | Q ® f e P The epithermal cross section is defined in terms of the resonance integral by the relation where u, is the lethargy at the lower energy bound of the epithermal group. The ratio Bg is obtained by equating the slowing down current from the & _ = _ _ epithermal group to neutrons absorbed or leaking in the thermal groupQ = ‘ 2] 4 zr,e Pe = za,t + DB ) 2 where Zr . is the "removal" cross section. > Let I, ¢ zT,e/“e where T denotes "total." Ignoring leakage gives fg =Zazt uR Py ' zt,e Combining these results gives. Op o F K(RI)i] [ o, =0 [Ut,s + K(RI)S] 1 s . . , ¢ B 217 where K EZT | The spectral index, K, is computed by Modric from input data. The reference element was the standard absorber referred to above. The product of the reference element cross section and the effective flux integrated over the core is computed from Modric output by dividing the fraction of neutrons as absorbed by the reference element by its atomic density, NS. This product is the desired number for use in the ERC calculations; the working equation becomes A [at o+ K(RI)i] G, V= — =—2 NS ["t,s + K(RI)S] i The thermal cross sections o, are computed from the 2200 m/s cross sections by multiplying by a factor that averages them over a Maxwell- Boltzman spectrum around the reactor temperature. This factor was computed for a l/v energy dependence of the cross section and applied uniformly to all fission product isotopes, except noble gases. In all, 115 fission product isotopes were so treated, linked by trans- mutation and decay ihto chains. Provision was mede in ERC-11 to remove each at a rate determined by its chemical or physical properties in rela- tion to the processing method. For instance, xenon is removed rapidly by transplratlon in an expansion chamber, whereas rare earths are removed only by dlscard of the fuel salt w1th 8 perlod measured in hundreds of days. ~ Xenon was treated separately in the Modric calculatlon, not only to determine accuratelyrlts effective cross section in the MSCR neutron spec- trum but also to ?ermit,special treatment of its exceptional behavior. It may be possible to remove xenon rapidly from the fuel solution by circulat- ing a portion of the salt through the dome of the expansion tank mounted over the core (Sec. 4 2 2) prov1ded the xenon does not diffuse rapldly into the moderator graphite. PTOVlSlon was made in ERC-11 to calculate the extreme cases (complete absorption in graphite vs. zero absorption) as well as intermediate situations where removal competes with absorption. 218 Samarium was also treated separately because of its impbrtaflce. The fission productcalculation'ig thought to result in a reasonably good approximation of the poisoning in reactors where the fission products are exposed to neutrons for a long time. The ingrowth of second and higher generation isotopes)by transmutation and decay is treated in detail. The transient period following start-up of a clean reactor with an initial loading of 235y is ignored, and all concentrations are calculated at their maximum, equilibrium values. Hence, the poisoning is overestimated some- what, thus providing a margin of safety in respect to assignmenfhof the cross sections and resonance integrals. o _ The fission-product reaction rate coefficient is qbtained'by reference to a specified standard absorber: FP FP o : o't = dR-ji— , ' (3) o where FP - - » » - C - = fission-product reaction rate coefficient, Ly reference material reaction rate coefficient, effective fission-product absorption cross gection, effective reference material absorption cross section. Q " The effective cross section ratio is obtained from a two-group formu- lation: o' F (ol ¢1/¢2 + 0, " ) (01 ¢1/¢2 * 62)R )FP ) (4) where o, = fast absorption cross section L th o (u) du (RI)a‘ = = , Yn Yen a} 219 Qj 1 absorption cross section averaged over the thermal flux, 2 g, = fast flux, - ¢2 = average thermal flux. For a two-group treatment, all neutrons removed from the fast group must either be absorbed or leak from the reactor while thermal: — 2 B 8, = I, ¢2+D3 g, R. ™1 1 2 Ignoring leakage, B L, — = 2, (5) %2 R 1 Also, — 2200 o, =f o (6) where [%)( T%?%fi) for a Maxwell- H 1l thermal spectrum factor = Boltzmann distribution, 02200 = 2200 m/s absorption cross section. Substituting Egs. (4), (5), and (6) into Eq. (3) gives - oy 2200]FP FP . R [K(RI)+-°a ] Co=C (7) | | o where g o2 Zay "L £’ R, - o= [R(RD) +02200)R L - K is calculated as follows: K = WS, 220 1 . . . . W = Fra s input to linkage section of input, w0 i 2 /ZR = value automatically calculated by MODRIC. 2 ™ The value of ({ must be specified if fission product 6ption 1 is specified on card BN=5. This value should be calculated using an estimated value for K. If fission product option 2 is specified on card BN-5, 'is cal- culated by the code using the latest value of K. The required values of 02200 and RI for a special reference material are built into the code for fission product option 2. The nuclear constants for arl/v absorber with a 2200 m/s cross section of 1.0 barns are built into the code. Therefore, to use fission product option 2, the referencé element must correspond to an artificial element in MODRIC which has cross sections for a 1l/v absorber with 02200 = 1.0. References l. J. Replogle, MODRIC — A One-Dimensional Neutron Diffusion Code, USAEC Report K-1520, Oak Ridge Gaseous Diffusion Plant, September 1962. 2. C. L. Davis, J. M, Bookston, and B, E. Smith, GNU-II, A Multigroup One-Dimensional Diffusion Program for the IBM-704, GMR-101, General Motors, November 12, 1957. 3. L. G. Alexander, ERC-5 Program for Computing the Equilibrium States of Two-Region Thorium Breeder Reactors, USAEC Report ORNL CF-60-10-87, Oak Ridge National Laboratory, October 1960, L 221 Appendix E FISSION PRODUCT NUCLEAR DATA® L. G. Alexander BCompiled from Appendix E list of references. Atoms per fission. Fission Yieldb 8, Deca Cross Resonance Number Isotope ? 2y Section Integral 233y 233y Ay sec g0, barns RI, barns 26 82ge 0.007 0.0028 2.1 1.4 27 81pp 0.0045 0.001% 3.3 60 28 82¢y 45 45 = 29 83gy 0.012 0.00544 205 201 30 84y 0.019 0.010 0.16 5.5 31 85Ky 0.006 0.00293 0.214 x 10-8 7 29 ) 32 86y 0.032 0.0202 0.06 0.04 ) 33 85Rp 0.019 0.010 0.91 0.67 34 87Rb 0.040 0.249 0.13 0.21 35 86gr 1.3 (0.6)° 36 875y 37 88gy 0.050 0.0357 0.0055 0.06 38 89y 0.065 0.0479 1.31 0.78 39 20y 0.065 0.0577 0.298 x 10-° 1 1.8 40 2lgr -~ 0.065 0.058% 1.2 9 41 927y 0.067 0.0603 0.15 0.55 42 9Bzyr 0.070 0.0645 1.1 28 43 gy 0.068 0.0640 0.076 0.2 iy Syr 0.057 0.0633 0.053 0.07 45 ?5Wb 0.229 x 107 46 25Mo 0.062 0.0627 13.9 109 47 26Mo 1.2 34 48 7Mo 0.053 0.0609 2.2 16 49 98Mo 0.052 0.0578 0.51 5.6 50 - 100M5 0.044 - 0.0630 0.3 6.2 51 99Te 0.048 0.0606 L 22.2 140 52 100gy, o 1.7 7 | 53 101gy 0.030 0.050 5 77 = 54 102gy 0,024 - 0.041 1.44 11 55 104py - 0.0097 0.018 0.7 8 56 93mn - - 0.016 0.030 18/ 1030 57 =~ 104pg. - . o 6 19 58 .105p3 0.005 0.009 11 76 59 106pg 0.0028 0.0038 6 12 60. - 107pg 0.0015 0.0019 10 40 61 " 108pg 0.0006 0.0007 10.7 169 62 . 110pg 0.0003 0.00024 0.28 10 ‘Efi; “Values in parentheses estimated by comparison with similar nuclides. 222 Fission Product Nuclear Data (continued) Fission Yield D Cross Resonance Number Isotope ‘ B, ecgg Section Integrsel 233y 235y A, sec go, barns RI, barnms 63 109 0.0004 0.0003 91 1420 64 1llcg 0.00025 0.00019 2 52 65 11204 0.0002 0.0001 1 13 66 11309 0.0002 0.0001 59,500 652 67 11404 0.0002 0.0001 1.2 15 68 11517 0.0002 0.0001 - 228 3300 69 1l6gp 0.006 70 125q¢ 1.56 (0.8) 71 126mpe 0.0024 0.0005 0.8 12 72 128p¢ 0,010 - 0.0037 0.3 2 73 1271 0.0039 0.0013 6.2 154 7, 1297 0.02 0.009 27 39 75 128ye — 5 45 76 129%%e 45 302 77 130ye 5 45 78 131xe 0.037 0.0293 120 806 79 132y 0.051 0.0438 0.2 1.8 80 133%e 0.152 X 10~° 190 1270 81 134%e 0.066 0.0806,, 0.2 . 0.6 82 135%e 0.067¢ 0.0641 0.211 X 10™% 3.344 X 10 0.6512 X 10 83 136xe 0.069 0.0646 0.296 x 10-3 0.15 0.1 82 1330g 0.062 0.0659 28 420 85 13405 0.110 x 10~7 137 1400 86 13505 - 0.067° 0.0641 8.7 62.0 87 13604 0.617 X 1076 88 1370g 0.072 0.0615 0.666 X 10~° 0.11 0.3 89 13804 0.362 x 10-3 8.7 62 90 134pg 2 (1) 91 136p4 0.4 (0.2) 92 137Bg 4.9 (2.5) 93 138p, 0.068 0.0574 0.68 0.3 9/, 1391 0.064 0.0655 8.9 11.0 95 1400¢ 0.061 0.0644 0.66 0.5 96 141¢c 0.251 X 10~ 97 14200 0.057 0.0595 0.94 1.3 98 . 143 0.601 X 10~5 6 (3) 99 141pn 0.059 0.064 11.5 23.5 100 142pp 18.0 (9.0) 101 143py 89 (0.45 x 10) 102 144 py 0.660 x 103 , 103 143Nq 0.052 0.0598 308 130 104 144Ng 0.041 0.0567 5 12 105 1458d 0.030 0.0395 67 245 106 146N 0.023 0.0307 10 25 107 . Y47§g 0.710 x 10~6 180 2510 dIncludes indirect yield from 13°I. ®Included in Xe yields. e 223 Fission Product Nuclear Data (continued) Fission Yield 8, Deca Cross Resonance Number Isotope ? ,{ Section Integral 233y 235y A, sec go, barns RI, barns 108 1483 0.012 0.0170 3.4 48 109 14954 | 0.963 X 10~% 110 150§g 0.0048 0.0067 1.5 14 111 151ng 0.700 x 10-2 112 147pm 0.017 0.0238 0.846 x 1078 180 2510 113 L48py 0.151 x 10™% 27,000 114 147gm 87 690 115 148gn . 9 50 116 149 0.0062 0.0113 87,770 2440 117 150gm 85 460 118 5lsm 0.0026 0.0045 0.301 X 10~? 10,260 3565 119 152gm 0.0017 0.00285 194 2500 120 153gm ' 0.410 X 10~ 121 154gm 0.00037 0.00077 5 25 122 153w, 0.00095 0.0015 382 1380 123 154Ey 0.137 x 1078 1500 750 124 155y 0.129 x 10~7 8490 4245 125 156y, 0.521 x 10~ 126 154G_d : 127 15364 0.00015 0.0003 58,000 1630 128 1564 0.00005 0.00013 4 by 129 15764 0.5 X 107% 0.78 X 104 0.24 X 108 740 130 15834 0.1 X 10-% 0.2 x 10-4 3.9 29 131 155G4 0.107 x 1074 132 159, 0.5 x 103 0.1 x 1074 46 420 133 89y | 0.148 X 10-° 130 (65) 134 07y 1.5 0.7 135 13571 (£) (£) 0.289 x 1074 136 135p, 5.6 (3) 137 151gy 8400 (4200) 138 1525y 5500 (2750) 139 160y , , 525 (262) 140 1307¢ 0.027 - 0.020 0.5 2.6 TOTAL 2.02285 2.06485 TYield of 1351 combined with that of 135Xe. e e i i+ a1 im o - ol 24 References 1. J. 0. Blomeke and M, F. Todd, Uranium-235 Fission-Product Production As a Function of the Thermal Neutron Flux, Irradiation Time, and Decay Time. II: Summations of Individual Chains, Elements, and the Rare-Gas and Rare-Earth Groups, USAEC Report ORNL-2127, Pt. II, Vols. 1, 2, and 3, Oak Ridge National Laboratory, November 1958. 2. J. D. Garrison and B. W. Roos, Fission Product.Capture Cross Sections, Nucl. Sei. Eng., 12: 115 (January 1962). 5. E., A, Neghew, Thermal and Resonance Absorption Cross Section of the _233U, 23 U, and 23%, Fission Products, USAEC Report ORNL-2869, Oak Ridge National Laboratory, March 1960. L, J. J. Pattenden, Fission Product Poisoning Data, USAEC Report ORNL 2778, Oak Ridge National Laboratory, October 1959. 5. C. R. Greenhow and E. C. Hensen, Thermal and Resonance Fission-Product Polsoning for 235U System, USAEC Report KAPL-2172, Knolls Atomic Power Laboratory, October 1961. 6. N. F. Wikner and S. Jaye, Energy-Dependent and Spectrum-Averaged Thermal Cross Sections for the Heavy Elements and Fission Products for Various Temperatures and C: °>°U Atom Ratios, USAEC Report GA-2113, General Atomic, June 1961. 7. R. L. Ferguson and G. D, O'Kelley, A Survey and Evaluation of 233U Fission Yield Data, USAEC Report ORNL CF-62-3-T71, Oak Ridge National Laboratory, March 1962. 8. E. C. Hensen, A Critical Examination of the Uncertainties in Predicted Gross Fission Product Poisoning, USAEC Report KAPL-M-ECH-8, Knolls Atomic Power Laboratory, March 1962, 9. E. C. Hensen and C. R. Greenhow, An Improved Generalized Analysis of Fission Product Poisoning and Thermal and Resonance Fission Fragment Cross Sections, USAEC Report KAPL-M-ECH-T7, Knolls Atomic Power Laboratory, September 1960. 10. C. H. Wescott, Effective Cross Section Values for Well-Moderated Thermal Reactor Spectra, Canadian Report CRRP-787 (Rev. August 1958); AECL-670. 11, J. 0. Blomeke, Nuclear Properties of 2>°U Fission Products, USAEC Report ORNL-1783, Oak Ridge National Laboratory, April 1957. O » 225 Appendix F TREATMENT OF DEILAYED NEUTRONS T. W. Kerlin Summarz Circulating fuel reactors lose neutrons because some of the delayed neutrons are emitted outside of the core. These losses depend on core residence time, external loop residence time, and decay characteristics of the precursors. A symbolic representation of the system is 7\1Nmfcvc 7\1NlEfEVE Py T N/ ] Heat Core Exchanger aiVZf¢chc where .-Aij = decay constant of the 1t precursor from fissionable material J’ B _ _ _ 513 = number ofilth precursors formed per fission neutron from fis- - sionable material j,- : | Ni‘c =_atoms of 1th precursor per ‘unit volume of fuel stream in the ,J core resulting from fissions in material j, : Ni’E = atoms of-lth,precursorrper.unlt volume of fuel stream in the J external loops resulting from fissions in material j, f = volume fraction of fuel in the core, 226 fE = volume fraction of fuel in the external loops, Vc = core volume, VE = exfernél loop volume, vjzfj¢fcvc = rate of productlon of fission neutrons in the core from fissionable material j. The precursor concentrations are described by these equations: dN, . ' ' - —_—C - L | dtc 613 3 f3¢ 7\N:v.jc ? ' (1) dNi'E \ | - . T - ijE ? (2) - E where tc = time in the core, tE = time in the external loops. The boundary conditions are: Ny 50(T) = My 450 (3) where - TC = time for the fuel stream to pass through the core, TE = time for the fuel stream to pass through the external loops. The solution to Egs. (1) and (2) are: B, WD 8 =\, .t N =—1J-J—fl—(1—e 1 c)+Nijc(O)e e (5) - B - ije LU " 227 ij°E NijE lJE(O) e . - (6) Note that the precursor production rate is assumed constant for the fuel stream during its stay in the core. This idealized case would exist only for uniform power density along the fuel stream or for core residence times which are short compared to the half-life of the precursor. The boundary conditions become BisV"e5” A 57, MgTe Kij - 11 —e Nljc(o) e = NijE(D) , (7) iR Nijc(o) = lJE(0) e - (8) Eliminating Nijc(o) in Egs. (7) and (8) and substituting the result in Eq. (6) gives ' ~A; 4T ) -A; 90%) is formed by the decay of 35T which has a half-life of 6.7 hours. The iodine remains in solution as the iodide ion (I'). Thus, at equilibrium, the rate of formation of xenon in the fuel isApr0p0rtional to the sum of the direct fission yields of xenon and iodine, here taken to be 0.066 atoms per fission. Al] of the xenon is released in the fuel salt, and since the half- life for the decay of 12°I is long compared to the time required for the fuel to make one complete trip around the fuel circuit (about 15 seconds), the rate of release of 12°Xe is nearly uniform throughout the fuel volume, being augmented somewhat in the core by the direct fission yield. For purposes of this study, the concentration of xenon in the salt was assumed to be uniform. The solubility of noble gases in molten salts is low (1), especially in mixtures of LiF-BeFz. A concentration of only 2 X 10!’ atoms of xenon per cc of salt at 1200°F is in equiliflrium with a partial pressure of xenon in the gas phase of one atmosphere. In the reference design case, the equilibrium pressure is about 0.06 atmosphere. Xenon thus tends to leave the salt at any phase boundary. It may form microbubbles clinging to the surface of the graphite moderator. It will tend to diffuse into the pores of the graphite. It can be removed rapidly by spraying a portion of the circulating stream into a space filled with helium or by subsurface sparging with helium. Xenon is also removed from the system by decay to 12308 and by re- action with neutrons to form !36Xe, which is stable and has a low neutron capture cross section. g =T 231 ~ Analysis Watson and Evans! have analyzed the equilibrium xenon poisoning re- sulting from the interaction of all these modes of production and removal, obtaining an equation which may be rendered in the form where P‘F' = | < v ¢ a. | d ¢ca fe + ] “ Y v P a+an Xe P.F. = — f £ (1) v v n n ¢ a-.-f;g- +,_§,+___d,-_ ¢ iV n n - f f £ . Poison Fraction, or number of neutrons absorbed by xenon per neutron absorbed in fissile isotopes, neutron productions, number of neutrons produced from all sources per neutron absorbed in fissile isotopes (2.21 in reference design), sum of fission yields of '3°Xe and 3°I, taken as 0.066 atoms per fission, neutron yield, or number of neutrons produced from fission per fission ( 2.50 in reference design), mean effective flux in reactor core (neut-cm/cm3 sec), - L RN ' — Captures effective 135Xe neutron capture cross sectlon neut cm Atom Xe, ce volume of fuel stream in core (600 fts), ‘volume of fuel stream (2500 ££3), decay constant for 135%e, 2.09 X 10 "3 /sec, rate of dlffu51on ©of xenon into graphite, atoms/sec, rate of removal of xenon by sparging, atoms/sec, number of atoms of xenon dlssolved in fuel salt- 232 The produce ¢c a is readily evaluated by reference to a multigrdup calculation. Thus P.aN V =FvA where Nc = concentration of xenon atoms in the core, atoms/ce, Vc = volume of core, cc, F = fission rate in core, fissions/sec, A = fraction of all neutrons captured in 135Xe. The fission rate is readily calculated from the power, using the conver- sion: 3.1 x 10%6 fissions/Mw-sec. Solving for the product 3.1x 102¢ Py [A | ¢C Q= vc fi) ° 7 (2) The ratio A/Nc is, at low concentrations, independent of N, and may be determined by means of a multigroup neutron calculation. In the ref- erence reactor it has a value of 2.22 X 10716, By insertion of numerical values, it is found that ¢c o equals 2.54 X Captures{cc 10™* atom Xe sec . ce The ratio fis/nf is equal to the volumetric rate of sparging divided by the volume of the fuel stream. Let Q be the fuel stream rate of flow through the core (160 ft3/sec) and r be the fraction of this diverted through a sparge or spray chamber. Then r'ls/nf = rQ/Vf . For the term n., Watson and Evans give a relation which may be rendered d / ng = M & [e D(F, a + 7‘)]1 2 o (3) 9 ¥ 233 where N* = gas-phase xenon concentration: that is in equilibrium with xenon € gdissolved in the salt, atoms/cc, A = area of interface between salt and moderator graghite, (For 810 & logs 8 in. diam X 20 ft long, A.’g = 31.5 X 10% cm®), e = porosity of graphite or fraction of graphite volume accessible to xenon, D = coefficient of diffusion of xenon in graphite, cm?/sec. The value of Nz is related by Henry s law to the concentration in the salt. - * N, = Ng KRT (4) where N, = concentration of xenon in the salt, atoms/cc, K = Henry's law constant for xenon in salt, 3.2 X 10-° moles Xe cc of salt, atom s R = gas constant, 82 cc-atom/mole °K, T = absolute temperature, 922°K, Values of K for xenon dissolved in various salts at various tempera- tures are given by Watson and Evans.! For MSCR salt at 1200°F, K is about 3.2 x 10 ? moles Xe/cc'saltfatbm.r Noting that n, = Nf Vf, one has, from | Eq'_'(3): . | | _' | _ A .‘[e D(¢ o + 7\)]1/2 , h./n, = & < ‘ | - (5) da' Tt o KRTV.f N -§UbStituting these results and numerical values into Eg. (1) yields 1 +42.6 X 10* (e D)2/2 Xe P.F. = 0.0584 . | 1.32 + 1000r + 46.0 X 10% (e D)1/2 234 Values of the poison fraction calculated by means of this equation for various values of r and the product e D are displayed below in Table G.1. Table G.1. Xenon Poison Fraction in MSCR Poison Fraction r . eD*=w eD*=10% eD*=10% eeD*-=0 0.1 0.054 0.0445 0.0173 0.0006 0.0L 0.054 0.053 0.0445 . 0.0052 0.001 0.054 0.054 . 0.053 0.0251 a In cm?/sec. Reference 1. G. M. Watson and R. B. Evans III, Xenon'Diffusion in Graphite: Effects of Xenon Absorption in Molten Salt Reactors Contalning Graphite, USAEC Report ORNL CF-61-2-59, Oak Ridge National Laboratory, February 1961. o 9 235 Appendix H THE EQUILIBRIUM STATE AS A BASIS FOR ECONOMIC EVALUATION OF THORIUM'REACTORS T. W. Xerlin / Introduction The equilibrium condition is currently being used as the basis for fluid~fuel reactor economiéfévaluatidn and for new computer code develop- ment. Because of this increasing applicatibn of calculations based on the equilibrium state, it is advisable to clearly define equilibrium and to assess the validity-df evaluatiéns based on the equilibrium condition. These problems wére considered in this study for thorium-fueled reactors, fueled initially with ?35U, and particularly for the molten-salt converter reactor (MSCR). The equilibrium condition is defined as that condition in which the reactor composition is time independent because of a balance between nuclide production rates and loss rates. It is important to note that these are the total production and loss rates (including feed, recycle, dischargé,.processing losées, ete. ) and are hot restficted to nuclear transformation rates. The mathematical formulation of the equilibrium state is obtained by setting the time derivative of the nuclide concen- tration equal to zero: o e m e et D e e ee oo ——_Qi+Ri+Fi+Ti+pi_ N(ti+di_+qi)_0,- | (1) dt T : | , | © vhere - N, = nuclide concentratién'of ;B nuclide, t = time, Q. = feed rate, . L . R. = recycle rate, e 236 F. = fission fragment formation rate, ;Ti = growth rate due to neufron captufé in other materials, Di = growth réte due to radidactifie-decay in»éther materials, ti = loss rate due to neutron capture in ith~nuclide, | di = loss rate due to radioactive decay of ith nuclide, qi'= processing loss rate. Equation (1) should be valid for reactor evaluation if the nuclide con- centration is near equilibrium (90-95%) over a large fraction (90-95%) of the reactor's operating life. - Estimates of the saturation behavior of the nuclides of interest in thorium-fueled reactors were made using the methods discussed below. Methods The time-dependent behavior of the nuclide concentrations in a thorium- fueled reactor, fueled initially with 235U, was calculated by solving Eq. (1) without the restriction that dNi/dt = 0. The treatment for the individual isotopes along with special assumption for each case is given below: 1. Fission Products Assume that fission products are produced at a constant rate, S, and are removed by neutron capture, radiocactive decay, and processing. This leads to dNi : 1 —= =8, =N, |[t, +a, +-]|, - (2) dat i il'i i T where T = time to process a complete reactor volume. Solving Eq. (2) gives 0 ® 237 5, —(%i +d. + ;rl-)t _..__._____...._l l - ) 1 . . » (3) t. +4d. + = ' 1 1 T N, (t) = The fractional saturation is 1 1\7.(1-,):—_1_-3(ti+di+T o ) A lower limit on the fractional saturation at any time may be obtained by setting ti = di = 0, This gives N, (t) "% =1-—e .. - (5) N (<) Neglecting processing losses and captures in 233Pa, the concentration of 233U is given by dN23. = = NO2 022 g — N°3 0:3 B (6) where 'N?3 = concentration of 233U, = o N it concentration of ?32Th (assuméd invariant), 092 = effective capture cross section of 232Th, # = neutron flux (defined below), 023 = effective absorption cross section of 233U. The neutron flux is given byi_rr Vg = P x 3.1% 1010, | (7) 238 | P 3.1 x 10*°. : f °f where P = reactor power (watts), V = fuel volume (cc), N, = concentration of fuel (atams/cc), - Op = effective fission cross section (average over all fissile nuclides). The term, VNf, is the total number of fissile atoms. Taking the average atomic weight of the fissile nuclides as 235 gives VN, = W X 2.563 X 1021 , (9) where W = mass of fissile material (grams). Substituting Eq. (9) into Eq. (8) gives | P| . 1.21 X 10712 @ = (V-I) X ———%.——'—"" (10) f Solving Eq. (6) and using Eq. (10) gives ] . P 033 | | | \ s 2s =X X 3.816 X 1074 T | M.t_l =)l —e £ ' , (11) N2 (w) where T = time (years). The term, P/W, is the specific power of the fuel in units of watts/g. The — ratio, 023/0f,'is about 1.2 for the molten-salt converter reactor (MSCR). " %) ¥ o - 239. 3. Uranium~234 The concentration of 234U is given by dl;? - N23 023 5 — N2 0@4 s, (12) where N?4 = concentration of 34U, 023 = capture cross section of 223U, Q n 24 absorption cross section of 23415, Solving Egs. (6), (10), and (12) simultaneously gives P 024 _——X -2 x 3.816 X 1074 T 24 W g F—(...t-l=l—e f N24(m) 23 24 Po Po —— 2 % 3.816X107% T —~=-2_x3.816X 1074 T . W Gf - e W Gf ( ) - . : ‘ - (13 l — — g24 a. The ratio, 024/°f’ is aboutro.érfqr_the,MBCR. 4. Uranium-235 The concentration of 225U at any time is that required by the criti- cality condition. For:these}estimétes; it is asgumed'that the fissile in- ventory is constant. This gives N25(t) = N25(w)'--+N253(°°)- W3 (s), Nififl=l+1“_zflfll(l_fi(_tl), (14) ‘N25(m) N25(m) st(w) 240 If 023 = 025, as is approximately the case in the MSCR, then a N23 () CR | = , (15) N2°(w) 1 — CR. where CR = conversion ratio (assumed constant). Using Egs. (11) and (15) in Eq. (14) gives 23 | Ja_« 3.816 x 1074 T N?%(t) ~CR Wo. TN S =1 +——e £ . . | (16) N25 (o) 1 -CR The assumptions of constant fissile inventory and constant conversion ratio are very crude, but will suffice for the qualitative evaluation desired in this study. 5. Uranium-236 The concentration of ?36U is given by dng’ = N?3 023 g —N%€ 02° g , - (17) where N26 = concentration of 236y, 025 = capture cross section of 239U, 026-= absorption cross section of 236U. Solving Eq. (17) and using Egs. (10) and (16) gives n) 241 26 Po | — =& % 3,816 X 104 T 26 ' N'(.tz-—l-'—e Waf . P oa i -4 Ga -4 —~———X 3.816 X107* T = = ——=x 3.816X1074T e ‘ — e " T-CR | 423 - (8 1 — 2 0.26 &a. The ratio, ogs/of, is about 0.1 for the MSCR. C o 6. Uranium-238 The 238U appears only in the feed along with 235U dN.zs N28 = (239U feed rate)|— — N?8 028_¢ , (19) dt N25 : : feed where N?8 = concentration of 238y, GZB = absorption cross section of 238y. The 237U feed rate is equal to the nuclear transmutation rate 2?5U_féed rate =N?5(t)0:5:¢_=' NZ?(#Q + N23(®) - N22(t) 025 g . . (20) ») UGS ARl by i b a EELE 105y ey AR PR 8 b 1 T 242 Solving Eq. (20) gives P oy y B st(t) —'I;G——X 3.816 x 10 T 1 528 —_——=c1l-e £ 1 -— = N2 () | - 1-CR \o2’ P U23 —— 2 x 3.816 X 1074 T ! W Op 1 1 1l — CR) e - 1 — . (21) 1 - CR} ¢23 CR - [o23 | - & | L 0-28 -1 . 0.28 a . a The ratio, 02%/0,, is about 0.1 for the MSCR. Results and Conclusions The results are shown in Figs. H.1l through H.6. They are discussed individually below: l. PFission Products Figure H.1l shows the saturation behavior of a material removed only by processing. In actual operation, the approach to saturation would be faster because of nuclear transformations. It is clear from Fig. 1 that the fission product nuclides which saturate slowly with respect to their nuclear reaction rates are at equilibrium (90% or greater) for 90% of a 30-year reactor life for processing times of less than 500 days. A 1000- day processing time (typical for the MSCR) gives equilibrium (90% or greater) for only 80% of the time. Therefore, the equilibrium treatment is doubtful for fission-product nuclides with low cross sections or long half lives. However, these effects are small. ) * N 1'0 N(t)/N(=) 243 ORNL-LR-DWG T = 10,000 days T = time required to continuous process a complete reactor volume of fuel. 6 8 10 12 14 16 18 20 22 24 26 28 30 _'Reactbr'operating Time (yéars). Fig. H.l. Fission Product Saturation. 1.0 N23(t)/N?? (=) 244 ORNL-LR-DWG. 72862 2000 /-—""' - 7 — 1000 / / 750 // l l e —— \ \ e S 6 g 10 12 4 16 18 20 22 Reactor Operating Time (years) Fig. H.2. 233y Saturation. 2 26 28 30 C O l.o N4 (t)/N2% () 245 ORNL-1R-DWG., 72863 \ \ s ——""‘——— \ v pad AN Ly [ [/ Yoo /1] 1 // (%) = specific power (watts/gram) © 10 12 14 16 18 20 22 24 26 28 30 Reactor Qperating Time (years) ‘Fig. H.3. 234U Saturation. = O st(t)/st(w) N23(t)/N?5 () N25(t)/N25 () o0 246 ) ORNL-IR-DWG. 72864 t%) = gpecific power (wattS/gram) CR=09 AN 1000 NBA 4 8 - 12 16 20 24 4 8 12 16 20 24 Reactor Operating Time (years) ‘Fig. H.4. 23°U Saturation. O | *" ¥} N26/N26(m) N26/N26(w) 247 N26/N26(¢D) ™ 2000 /// / 1000 ,;,,—,.——-;———;—____;____—e:::::::: 750 L " P _ ) e W " 200 S - 4 g 12 16 20 24, 28 Reactor Qpéra_ting Time (years) Fig. H.5. 236U saturation. = specific power (watts/gram) 8 12 16 20 Fig. H.6. 238y Saturation. 24 28 O 8 " .l 249 2. Uranium-233 Figure H.2 shows the saturation behavior of 233U. The curves show that equilibrium (90% or greater) exists for 90% of the reactor life only if the specific power is greatér than about 1800 w/gram. A specific power of 750 w/gram (typical for the MSCR) insures equilibrium (90% or greater) for 80% of the 30-year reactor life. 3. Uranium-234 Figure H.3 shows that 34U saturates very slowly for all practical values of specific power. A specific power of 750 w/gram (typical for the MSCR) gives an average concentration over the 30-year reactor life only 65% of the equilibrium value. 4. Uranium-235 Figure 4 shows the saturation behavior of ?2°U. For a conversion ratio of 0.8 and a specific power of 750 w/gram (typical MSCR conditions), the 223U concentration is within 10% of the equilibrium concentration for 65% of the reactor operating life. 5. Uranium-236 Figure 5 shows that the approach to equilibrifim is quite slow for 236y. TFor a conversion ratio of 0.8 and a specific powver of 750 w/gram (typical MSCR conditions), he 236U concentration never reaches 90% of the eguilibrium concentration. - 6. Uraniuma238 | Flgure H.6 shows the approach to saturation for 238U. For a con- . version ratio of O. 8 and a spec1f1c power of 750 w/gram (typical MSCR ”condltions), equllibrlum (90% or greater) is insured for 20% of a 30-year reactor operating lifetime. The results for the six materlals considered in this study are shown ~in Table H.1 for MSCR conditlons._ Consideration of the methods used to obtain the results in Table H.1 does not indicate high accuracy. However, the results should be qualitatively 250 Table H.1. Percentage of MSCR Lifetime Having Nuclide. Concentrations Within 10% of Equilibrium MSCR Lifetime Having Nuclide Concen- Nuclide | tration Within 10% of Equilibrium | @) o Fission Products - 80 (1000-day processing) 233U B | 80 234y D 20 235U | | 65 236y ' o 0 238U 20 correct; and they should create some concern over the validity of the equilibrium state as a suitable condition for reactor economic evaluation. Table H.2 shows the expectéd direction of the error introduced by assuming the equilibrium condition for the nuclides considered. It is apparent from Table H.2 that it is not possible to predict whether equilibrium cal- culations are intrinsically optimistic or conservative from these results. Also, since the relative magnitude of the competing effects depends on re- actor type, the characteristics of the particular reactor must enter into an assessment of the direction and magnitude of the error associated with the equilibrium assumption. The magnitude of the effect of these factors on reactorrecqnamic evaluations is not known. However, an extensive study should be made to examine these problems. Reactor evaluations based on the equiiibrium con- dition are so convenient, economical, and-unambiguous that they éhou;d be used if possible. 251 - } ey Table H.2. Effect of the Equilibrium Assumption on Calculated Reactor Performance Material - Effect on Calculated Performance Fission Products Conservative (overestimate) 233y Optimistic (overestimate) 234y Conservative* (overestimate) 235y Optimistic (underestimate) 236y Conservative* (overestimate) 238y ‘Conservative* (overestimate) » *These conclusions are based on an equilibrium - calculation with no corrections. These materials are actually calculated using adjustment factors which * average the concentrations over the life of the re- actor. The direction of the expected error under this assumption is not known. » Estimated physical propefties of three fuel salt mixtures at 1200°F and one coolant salt at 1062°F for use in heat transfer and pressure drop calculations of the MSCR stud& aré listed below in TablefI.lo 252 Appendix I ESTIMATES OF PHYSICAL PROPERTIES OF LITHIUM-BERYLLIUM MSCR FUEL AND COOLANT SALTS R. Van Winkle ‘Introduction Table I.1, MSCR Salt Properties Mixture MSCR No. 1 MSCR No. MSCR No. Coolant Temperature, OF 1200 1200 1200 1062 Composition mole % 71-16-13 68-23-9 66-29=5 66-34-0 LiF=-BeF,.-ThF 2 4 Liquidus Temp, °F 9y1 887 860 851 Mol. Wt. 66,03 56,2 46,2 33,14 Density, lb/ft3 215.6 190.1 163.0 120,5 Density, g/cc 3,454 3,045 2,610 1,931 Viscosity, lb/hr-ft 24,2 21 18,9 20,0 . Thermal Conductivity 2,67 2,91 3.10 3.5 Btu/hr-ft-°F Heat Capacity, 0.318 0.383 0.449 0,526 : Btu/1b-°F The bases for these estimates, some temperature-dependent relation- ships and atom number densities are given for cases of interest, £k L] w tions of the mixture LiF-BeF, -~UF listed in Table I.2 at temperatures of 600°, 253 Viscositz Experimental values of the'viscosity of several different composi- 2 ~ThF,, (and some estimated values) are 700° and 800°C, TheSe are plotted in Fig. I.1l which also includes plots of what appear to be rea- sonable estimates of the viscosity of MSCR Salts 2 and 3. The estimates on these two salts depend on the fact that their compositions lie between those of Mixture 75 and Mixture 133 (which is the same as MSCR No. 1); hence, their viscosity curves may be expected to lie between the curves of the two known mixtures. Viscosities and temperature-dependent vis- cosity equations of MSREIffiél and coolant salts are listed in Tables I.3 and I.4 for comparison, The viscosity equation for Mixture 133 (MSCR No. 1) is (1): n. = 0,0526 exp(4838/T °K) centipoise Table I.2, Some Physical Properties of Various Lithium-Beryllium Molten Fluoride Salts Mixture Composition, m/o Viscosity, cp Heat Capacity(2) 13,4 . Number LiF-BeF,-UFy-ThF, 600 700 800°C at 700°C Mol. Wt. 7% 69 31 0 0 7,5 4.9 3.45(5) 0.67%(5) 32,1 7.25 4,58 3,10(6) 75 67 30.5 2.5 0 8.4% 5,5% 3.8%(5) 0.57%(5) 39.5 11 71 16 1 12 13.0 7.1 #,8(5) 0.37%(5) 66,02 112 50 50 0 0 22,2 10,7 5.95(5)- 0.65%(5) 36,5 T T 2 e sloae) 131 60 36 0 13,0 7,96 5,30(6) -- 41,96 132 57 u3 0 13.4 7.38 4,50(6) - 35,03 13 71 16 0 13 7.55 4,76(1) 0.306(1) - 66,03 #Estimated values (all others listed are experimental). Numbers in parenthesis are references. 254 ORNL-LR-Dwg. 62611 TEMPERATURE,°F 9501000 11 1200 1300 1400 1500 16001700 1800 20 ©® ORNL-2150 X MIM=1086 ® Hoffmsn 15 ity . O O ~N O 0w O ViscostTy, Cp H o COMPOSITION (m/o) I1 Be U Th 2 69 3 wm o 67 3005 2.5 bt L 16 1 12 1.5 50 50 oo o L 16 o0 13 500 5§50 600 650 700 800 900 1000 TEMPERATURE, °C | 800 850 900 950 1000 1100 1200 1300 l I I I l | it {1 | TEMPERATURE, °K . Fig. I.1l. MSCR Salt Viscosity. A . oW *} L{] 255 Table I.3, Composition and Properties of Fuel Salt for MSRE I, II. I1I, Chemical Composition LiF BeF2 ZrFu ThFl+ UFu Molecular wt Physical Properties Density (above liquidus) 1b/ft t in °F @ 1200°F 3 Liquidus, °F Heat capacity, Btu/1b-°F Solid 212 - 806°F Liquid 887 - 1472°F @ 1200°F Heat of fusion (@ 842°F), Btu/lb Viscosity, cefitipoise, T in ©F Range: 1122-1472°F @ 1000F _Thermal condtétivity;fiBtu/hr-ft—°F ~tdin°F @ 1200°F Mole Percent 70 23 43,59 2 177.8 - (1,9% x 10 “)t - 154.5 8u2 0.132 + (4.033 x 10~ Ht 0,575 - (9.99 x 10'5)t 0.455 138.6 0.1534 6476/T 7.64 2.74% + (5.516 x 10~ ')t - (1.37 x 10~7)t2 4 3.21 . Table I.4, Composition and Pro : - Coolant Salt for MSRE perties of I. Chemicél'Composition,,,'___ LiF BeF2 ITI. Molecular Weight Molé Percent 66 3y 33,14 256 Table I.4, Continued - III. Physical Properties (at normal ave operating Temperature, (1062°F) Density, 1b/ft> 120,5 Viscosity, lb/ft-hr 20,0 Specific heat, Btu/1b-°F o Solid (122 - 680°F) 04210 + (8,71 x 107 Ot Liquid (896 - 1508°F) 0.17% + (3.31 x 10~ 1)t @ 1062°F 0,526 Thermal conductivity, 3.5 Btu-ft/ft2-hpr-°F Liquidus temperature, °F 851 Heat CaEacit! The temperature-dependent equation for heat capacity of Mixture 133 is (1) CP = 0,473 - 0,000238 T cal/g-°C (T = °C) Values of heat capacity of MSCR Salts 2 and 3 at a given temperature may be estimated by interpolating between published values of other salts of different molecular weights, since heat éapacity of molten salts probably varies inversely with molecular weight. Published values of heat capa- city at 700°C of some known salts are shown in Fig, 1.2, which contains a‘plot of heat capacity as a function of molecular weight, Heat capacity relationships for MSRE fuel and coolant salts are shown in Tables I.3 and I.4. Thermal Conductivity Like heat capacity, thermal conductivity can be expected to vary inversely with molecular weight. An estimate of the thermal conductivi- ties of the MSCR fuel salts has been made by extrapolating the published O o | A e bt abaenan ks cxmsanmb b 1o ot s bt g 18 b «} L)) 1 1] LY} %) Heat Capacity, Btu/lb:°F Thermal Conductivity, Btu/hr-ft-°F 257 ORNL-LR-DWG, 62612 “ | I _ MSRE Coolant - - -"“tc:;____ MSRE Fuel MSCR No. 3 S 3 -‘-"‘-JE— MSCR No. 2 - 2 0.7 ‘}(/“'NBRE Coolant Salt (700°C) 0.6 2 | MSRE Coolant Salt (572°C S .){F- ) ‘\\ | 0.5 >\ _ ™| - MSBE Fuel Salt N (700°C) ‘{-MBCR No. 3 N _ 0.4 MSCR No. 2 MSCR No. 1 (700°C) 0.3 30 40 : 50 60 70 Mol. Wt. Fig. I.2. MSCR Salt Thermal Conductivity and Heat Capacity. 258 values of the lower molecular weight MSRE salts to the higher weights of the MSCR salts. This extrapolation is shown in Fig. I.2, However, the published values may not have much precision or accuracy (;); Density Temperature dependent relationships for calculating the densities of the three MSCR salts are (2): -y 3 e -4 -4 PMSCR No. 3 2,993 - 5,9 x10 T The basis for these estimates is the same as given in reference (3), page 123 with an added term for the ionic volume of thorium equal to 2,82 - 2,94 x 107°T, Liquidus Temperature Figure I.3 (from reference 4) was used to obtain the liquidus tem- perature for the MSCR salts. Mixture 133 (MSCR No, 1) has a liquidus temperature of 505°C (p, 58, ref. ). g O A 259 ORNL-LR-DWG 37420AR2 Th, 4444 0 "o TEMPERATURE IN °C COMPOSITION IN mole %% - 1050 LiF-4ThE, s TLiF-6ThE; : % LiF-2ThE 3LiF-ThFy ss 4 1000 LiF-2The, . - . 950 P 8O7 2LIF'BGF2 900 7LiF-6ThE P 762 850 P 597 £ 568 3LIF-ThE oo\, >, 4 £ 565 %0 650 3. B O\ & ‘O, N N3 550 526 0 \ . , N/ 845 2LiF-BeF; 500?450, 400 400 450 500 548 - .P465 - £ 370 el Fig. I.3. The System LiF-BeFp-ThF,. b, 5. 260 References -~ ] Private communication from H., W. Hoffman, Private communication from S, Cantor. Oak Ridge National Laboratory, MSRP Semiann. Progr. Rept. August 1, 1960 to February 28, 1961, USAEC Report ORNL 3122, ' C, F. Weaver et al., Phase Equilibria in Molten Salt Breeder Reactor Fuels. 1. The System LiF-BeF,-UFy~ThFy, USAEC Report ORNL-2896, Oak Ridge National Laboratory, December 27, 1960. S. I, Cohen, W, D, Powers, and N, D. Greene, A Physical Property Summary for ANP Fluoride Mixtures, USAEC Report ORNL-2150, Oak Ridge National Laboratory, August 23, 1956, (Declassified 11/24/59.) B, C. Blanke et al., Density and Viscosity of Fused Mixtures of Lithium, Beryllium, and Uranium Fluorides, USAEC Report MLM-1086, Monsanto Research Corporation, March 23, 1959, W. L. Breazeale, Revisions to MSRE Design Data Sheets - Issue No. &4, MSR-61-100, August 15, 1961. ~ O AL = » 3 " 261 Appendix J FUEL AND CARRIER SALT COST BASES W, L. Carter Introduction o9 LiF, Ber, ZrPu and NaF to determine current market prices. The data are evaluated A survey was conducted among suppliers of ThFu, ThO and a recommended set of values to be used in molten salt converter and breeder reactor calculations is presented, The cost data are needed in calculating fuel cycle costs, The following values are recommended for use in molten salt reactor calculations: ~ ~ ThF, $ 6,50 per pound BeF2 7.00 per pound LiF 14,70 per pound ZrF# ' 4,00 per pound One of the purposes of the study of thorium breeder and converter reactors is to furnish comparative fuel cycle cost data on the various systems as well as nuclear performance data. The market survey was con- ducted to obtain current.and reliable price information on several chem- ical compounds which will be needed 1n rather large inventory and for which the consumptlon rate may -be s;gnlflcant., The inquiries were con- cerned prlmarlly with molten salt reactor materials, namely, thorium fluorlde llthlum florlde, berylllum fluorlde, zirconium fluoride and ,:sodium fluoride' in additlon, prices. were obtained for thorium oxide, : Several manufacturers of these chemlcals were contacted and a sum- '"mary of their prlce schedules 13 given in ‘Tables J.1 through J.6. Since it is approprlate to assoclate a date with a market quotatlon, it may be noted that these data were obtalned dur;ng the period November 1961 - January 1962, There is one exceptlon: the comparative figures quoted from document Y-1312 were published in December 1959, Table J.1. Cost Data for Thorium Fluoride c9¢ Cost $/1b ThF,* $/1b ThF, $/kg Th* $/kg Th Vendor or Source of Information o y-1312° " GCD ¥-1312 Quantity Initial order of 127,000 kg 6.00% 6.50 17,524 18.98 Tth, | Replacement rate of 37, 100 (d) (a) kg ThF,/yr Tnitial order of 1,271,000 6.00% 17.52¢ kg ThF, Replacement rate of 371,000 (d) (a) kg ThF,/yr a‘No assay given for the material. GCD- General Chemical Division, Allied Chemical Corporatlon, P.0. Box 70, Mbrrlstown, New Jersey. This vendor says price is only a rough estimate. ®R. G. Orrison, Thorium Metal Processes, Y-1312 (December 18, 1959) is not based on the indicated guantity. " This price @No dlstinctlon made in price because of quantity or rate. - 41 »n " v » i\ " Table J.2. Cost and Composition Data for Thorium Oxide Cost Data Cost $/1b ThO, $/1b ThO, $/1b ThO, 4$/kg Th $/kg Th $/kg Th Vendor or Source of Information AP® AP y-1312° AP AP Y-1312 Material Designation ~ Code 111 Code 112 Code 111 Code 112 Material Form | Powder Powder Not stated Powder Powder Not stated Cost quoted for Tnitial order of 127,000 kg - 7.00 7.50 5.75-8.50 17.52 18.78 14.39-21.28 ThOp L | Repiacement rate Of 37 lOO 7.00 - 7.50 17.52 18.78 kg TnOz/yr - N Initial order of 1,270, 000 - 7.00 7.50 17.52 18.78 W kg ThOg ' Replacement rate of 371,000 7.00 7.50 17.52 18.78 kg ThO, /yr ‘ Composition Data Typical Analysis {ppm unless indicated) Vendor or Source.of Information AP AP . Y-1312 Element or Compound ThO, - 99% min . 99% min Not given ®AP — American Potash and Chemical Corporation, 99 Park Avenue, New York 16, New York. °r. @. Orrison, Thorium Metal Processes, Y-1312 (December 18, 1959). This price is not based on the indicated quantity. Table J.2 (continued) Vendor or Source of Information Elememt or Compound (continued) Rare earth oxide Sulfate, SO3 Phosphate, P20s e Cal - MgO Na + K + Li Silica, Si02 Boron, B Uranium, U Loss on ignition Sm - Eu Ga Dy Composition Data Typical Analysis (ppm unless indicated) AP 50 100 50 50 100 100 2000 500 5000 10 10 AP 30 50 10 6 10 1 1l 5 0.1 10 5000 12 0.2 1 1 Y-1312 O 792 L3 $ w* . - ‘ L] Table J.3. Cost Data® for Lithium Fluoride Costb $/1b LiF $/kg Li fQuantlty Inltlal order of 16,000 kg Ii (c) . () Replacement rate of 5,100 kg Li/yr (c) (c) ¢9¢ ®The sources of information are Atomic Energy Commis- _ sion-Oak Ridge Operations, Oak Ridge Gaseous Diffusion Plant, and Union Carbide Company — Y-12 Plant. b'I'he price is for (c) at. % 14 and includes a basic charge for the material produced as the monohydrate (LiOH-H50) plus a conversion cost for LiOH-H,O0 — LiF plus a feed cost plus an AEC overhead cost. cClassified information. Table J.4. Cost and Composition Data for Beryllium Fluoride Cost Data o Cost $/1b BeFa $/kg Be Vendor BBCo® BCorpb BBCo BCorp Quantityc Tnitial order of 23,800 kg BeFz 6.66 7.25 76.52 83.29 Replacement rate of 6,950 kg BeFa2/yr 6.00 7.25 68.93 83.29 Initial order of 238,000 kg BeF, 6.48 6.95 T4 .45 79.85 Replacement rate of 69,500 kg BeFp/yr 6.00 1 6.95 . 68.93 79.85 Composition Data Manufacturing Typical Analysis of Element or Compound Specifications "~ Manufactured Material BeF 99.5 + 0.5 wt % | 99.5 wt % min Fe 400 ppm max 25 ppm Ni 100 ppm max v 20 ppm Ci:» 100 ppm max <1 ppm Al 200 ppm max 90 ppm S 500 ppm max ' 750 ppm %8rush Beryllium Company, 5209 Euclid Avenue, Cleveland 3, Ohio. bThe Beryllium Corporation, Reading, Pennsylvania. cShipped as 1 X 1 X 1l-inch lumps. The analysis given is for BBCo material; no analysis was given for BCorp material. O ' ‘ ‘ © 99¢ 9 » " ¢ w i\ ") ~Table J.5. Cost and Component Data for Zirconium Fluoride Cost Data Cost $/1b ZrF, $/kg Zr Vendor Tap® gcp® TAD | GCD Quantity Initial order of 55,100 kg ZrFy 4.00 4.50 16.13 18.15 Replacement rate of 16,100 kg ZrF,;/yr 4.00 (c) 16.13 Initial order of 551,000 kg ZrF, 3.55 (c) 14.32 Replacement rate of 161,000 kg ZrF,/yr 3.55 (c) 14.32 Composition Data .92 Typical Assay (wt %) Vendor TAD GCD Element or Compound ZrF, / Not given 98.5+ Chlorides 0.007 S ' _ 0.003 Hf ' 0.01 Fe 0.03 Ni _ 0.003 &TAD — Titanium Alloy Manufacturing Division, National Lead Co., 111 Broadway, New York 6, New York. No assay was given; however, the bid was for Hf-free material. bGCD — General Chemical Division, Allied Chemical Corp., P.O. Box 70, Morristown, New Jersey. This vendor says the price is approximate. ®No cost distinction is made for quantity or rate. Table J.6. Cost and Composition Data™ for Sodium Fluoride Cost Data | Costb $/1b NaF $/kg Na Shipping Containers 100 1b multiwall paper bags 0.135 | 0.542 375 1b leverpak fiber drums 0.139 0.558 Composition Data Typical Assay (wt %) Elements Na Major constituent Ca 0.05-0.5 Al - } ' 0.03-0.3 Si : . 0.01-0.1 Fe, Mg : _ 0.0005-0.005 each Cu Trace, <0.0001 Sc Not detected, <0.1 K, Ba Not detected, <0.01 each B, Mo Not detected, <0.001 each Ti, Mn, Co, Ni, V Not detected, <0.0005 each Cr, Ag Not detected, <0.0001 each 8711 information is from the Blockson Chemical Company, P.0. Box 1407, Joliet, Illinois. PPor an initial order of 75,800—758,000 kg NaF and a replacement rate of 22,100-221,000 kg NaF/yr. w o O ' _ 0 89¢ 269 Bases for Establishing Prices It was assumed that prices would be established for a large molten salt, power-producing system, Two conditions were visualized: (a) A 1000 Mwe (2500 Mwt) station and (bj ten of these 1000 Mwe stations in simultaneous operation. These conditions established the initial inven- tory requirements. The consumption rates were calculated by assuming that the fuel salt would be discarded after removal of fissile material on a 1000-day cycle, It was assumed that thorium and carrier salt could not be decontaminated and recovered. Vendors were asked to quote prices on the basis of producing mate- rials in the quantities desired by existing methods and according to current specifications. It was not considered appropriate to ask a ven- dor in an infbrmation~seekihg survey such as this to do much research into manufacturing procedures and schedules if his operations would be significantly affected by these additional quantities. Consequently, no rigid specifications were affixed to these chemicals other than the ob- vious one that all materials should have extremely low concentrations of high neutron cross section materials. General Comments on Price Quotations Thorium Fluoride Only one manufacturer was interested in making a quotation for ThF,, and it was admitted that this was a rough approximation. The price com- pared favorably with the value quoted by orrison (1) in a previous market survey. Zirconium Fluoride Vendors were asked to quote on hafnium-free Zrf,. The quantities requested are apparefitiy quife'large cbmpared with available production, 270 Beryllium Fluoride The suppliers indicate that there would be no problem in supplying the quantities requested. Beryllium fluoride is an intermediate compound in the production of befyllium metal. Although the vendors state that | the prices quoted are tentative, they are probably rather accurate, Sodium Fluoride This chemical is available in large supply. The quoted prices should be quite firm. Sodium fluoride is so inexpensive that its use in a reactor contributes negligibly to the fuel cycle cost, Thorium Oxide Thorium oxide is availablelin 1arge supply; the quoted prices are perhaps rather firm. The quotes compare favorably with the values given by Orrison (1). Lithium Fluoride Lithium fluoride occupies a singular position in the molten salt fuel system particularly with regard to availability and price. Since the lithium content must have a high ’Li assay, the only source of mate- rial is from AEC production facilities. AEC-ORO, Y-12 and ORGDP per- sonnel (2) were instrumental in developing a classified price schedule for high isotopic purity 7Li in quantities to.meet the requirements of a large molten salt power installation, Lithium is produced as LiOH+ A value of 0.05 is assumed for the fraction of total heat produced in the moderator, essentially the same as for the MSRE. Multiplying the heat rate by the overall net thermodynamic efficiency, one obtains the net power output. An assumed value of 40% efficiency is used. With these assumptions Eq. (4) becomes P Dc3 = 0.0196 = Mwe (net) . (5) l1-f r v o O e ak (3 277 Figure K.2 is a plot of Eq. (5) for the range of interest. Also shown in the figure are 500 and 1000 Mwe condition for fuel volume frac- tions of 0.10, 0.15, and 0.20. With the figure, and for a given set of conditions one is able to estimate qfiickly the maximum graphite size per- mitted by thermal stress considerations. For example, if it is desired to provide 1000 Mwe with single core of 20 ft diam and 10 vol. % fuel, it is seen from the figure that the largest graphite moderator prism is limited to about 9 in. across flats. It should be pointed out that the data used in constructing the figure are subject to considerable uncertainty. Design data for large sections of graphite such as likely to be used in the MSCR are, of course, unavailable at this time. However, it is believed that the results obtained from the figure will be conservative. 278 QRNL-LR-DWG 76828 r,* 20" 25" 3.0" 1?\° l/ . ] Eb " . .‘? 35 1800 = s & 1600 = 7 O > 4.0" 1400 < , / g NS 7/ / 1000 mwe, / 070 — — T T = 7/ Joocomws /|, 0.5 | ' 7 1200|——f———fZLIE e S ——— ’aoa’il-v-:-&'—’g:,g o o e—— e — D) A A Sw— 4.5- o z / ) / E. 1000 / / / 7 > o-l.: / / / / / / = 800 / Z - 500m|v;/fv.020/ / / S500mwe’, ,20.15, ' .‘_5.02’"7"!{.!1’.:0" — ..../.-_ e e —— 160" 600 - 79 z T 2 = _ 400 N W\ \ / '0. |o. LN \ Z oIO(')C) 2000 3000 4000 5C00 6000 7000 8000 Cube of Core Diam, ft. ] 3 1 i 10 12 15 I7.5 19 20 Core Diam, ft* Fig. K.2. Thermal Stress Power Limitation. O O ) 5 Ll kil e ot st sl s st 279 References Ralph Hankel, Stress Temperature Distributions, Nucleonics, Vol. 18, No. 11, pp. 168-169 (November 1960). B. L. Greenstreet, Oak Ridge National Laboratory, personal communica- tion. S. E. Moore, Oak Ridge National Laboratory, personal communication. L. G. Alexander et al., Thorium Breeder Reactor Evaluation, Part 1 — Fuel Yield and Fuel Cycle Costs in Five Thermal Breeders — Appendices, ORNL-CF-61-3-9, Appendices, p. 33 (May 24, 1961). C. W. Nestor, MSRE Preliminary Physics Report, ORNL-CF-61-4-62 (April 9, 1961). 280 Appendix L VOLUMES OF FUEL SALT AND INTERMEDIATE COOIANT SALT FOR ~ 1000 Mwe MOLTEN SALT CONVERTER REACTOR D. B. Janney Introduction The volumes of the salt streams (including heels in dump tanks, etc.) were calculated for the reference design described in Sec. 4. Fuel Salt Volume Table L.1. Volume of MSCR Fuel Salt A. Reactor - 1360 £t B. Piping 320 ft3 C. Pumps 130 ft3 D. Primary Heat Exchangers 575 £t E. Dump Tanks, Control Tanks 115 £t> TOTAL 2500 ft? These volumes were calculated as follows: A. Resactor Core (L = 20 ft, D = 20 ft) 0.785 (20)? x 104 = 630 ft? Annulus (L = 20 ft, t = 1 in.) | %fi-(20)2 = 105 £t3 Top Plenum (h = 13 in., D = 20 ft, incl. core hold-down grid* of hg =4 in., t = 1 in.) *See Fig. 4.8, O L) «) L] 281 o7 oy - 3] = (2220 0 (3] ] - 20 - 6 X 5.2 Bottom Plenum (h = 11 in., D incl. core support grid¥* of h t =1 in.) & ;;]'_ [0.785 20 2 [0'785 (20)% x 35 6 X 5.0 Dome (L = 6, D = 6) Q. REACTQR TOTAL Piping 2 1l x 24 (3] 3] - 288 — 36 785 (6)% X % (1iq. vol.) Pump Suction (6 ft, 14 in. Sch. 20) Pump Discharge (20 ft, 12 in. Sch. Reactor Inlet (25 ft, 10 in. Sch. Misc. Piping (Estimate) PIPING TOTAL Pump Bowl (8) (Equiv. ann. 1 ft, gquiv;'dépth 1 X o | 40) H 20) . 82.5 25X.']T4—4—'X8 1/2 £t) T X 1(31/2) x11/2x8 = 290 250 85 1360 50 125 115 30 320 130 £t ft3 ££3 £t3 £t3 i3 i3 ft3 2 282 D. Primary Heat Exchanger (8) (Shell ID 43.75 in., shell length 13 1/2 ft) 0.785 (43.75)2 _ 0.785 (0.5)2 4050 ' [ T4 T2 ] 13.5 x 8 21)° 1~ 23)] 5 i 3 + [0.667 X T 12) (; 10.4) 8 =530 + 45 = 575 £t Coolant Salt Volume Table L.2. Volume of Coolant Salt A. Superheaters 2425 ft3 B. Reheaters 485 ft> C. Primary Heat Exchangers 510 ft> D. Piping 2710 £t E. Pumps 100 ft3 F. Flush Salt 2385 £t° TOTAL g6l5 ft3 These volumes were calculated as follows: A. Superheaters (16) (Shell ID 31.5 in., U-shell length 58 ft, 0.5 in. tube bundle annulus) 3 salt vol. . ros ‘ (0.0029 pt? S22 YO ¢ 785 x 58 X 16) f [TX 25 GL) 58 x 16] = 2112 + 513 = 2425 247 O 9 ») 283 Reheaters (8) (Shell, ID 31 in., U-shell length 23.6 ft, 0.5 in. to be bundle annulus) 3 salt vol. (0.0029 £t? S X 766 X 23.6 X . [w X 0.5 (30.5) X 23.6 X 8 144, Primary Heat Exchangers (8) (Tube ID 0.43 in., tube length 25 ft, 2025 tubes) g ] = 424 + 61 = 485 £t3 2 [2:785 L0:43)" » 25 x 2025 x 8] + 200 (H-X neads) 144 = 410 + 100 = 510 ft3 ‘Piping (avg. lengths) Primary H-X outlet (155 ft, 14 in. Sch. 20) 140.5 _ 3 155 =355= X 8 = 1250 £t Primary H-X inlet (125 ft, 14 in. Sch. 20) 125 57— % 8 = 970 ft Superheater inlet/outlet (5 ft, 8 in. Sch. 20) | 5258 35 . 60 £t? 144 "y Reheater inlet/outlet (5 ft, 5 in. Sch. 40) Pump Suction (70 ft, 12 in. Sch. 20) 'PIPING TOTAL Pumps (System pressurizingrand salt sampling volume only) Flush Salt (Amount required equal to reactor system volume) 284 . ] L 5977 X 16 = 10 ft 70 118 8 = 460 £t° 14 - 2710 £t3 100 £i3 - 2385 £t3 o &£$ L4 285 Appendix M EVALUATION OF A GRAPHITE REFLECTOR FOR THE MOLTEN SALT CONVERTER REACTOR T. W. Kerlin " Introduction Nuclear calculations on large molten salt converter reactors (MSCR) indicate that the neutron leakége is large enough (2 to 4% of the neutrons produced) to warrant consideration of a graphite reflector. A reactor was chosen from a set presently under study to evaluate the desirability of a reflector. This reactor, which current results indicate is near the optimum with respect to fuel cycle costs, has the characteristics given in Table M.1. Table M.l. Typical Characteristics of MSCR Diameter of core, ft 17.7 Height of core, ft 17.7 Carbon-to-thorium ratio o 293 Fuel salt composition 68L1.F-23BeF2-5ThFl+ Fuel salt volume'fractiog 0.10 Fuel processing rate, ft“/day 2 The core con31sts of a graphxte matrix inside a 2-1n.-th1ck INOR-8 vessel. Because of the different coefficients of thermal expansion, the vessel will move away_from the graphite when the reactor is at power, “creating én annulus. Since no adequate metal-to-graphite seal is avail- able, this annulus will contain fuel salt., The designer may choose to place a reflector between the core and the annulus by merely increasmng the size of the graphlte region and omitting _fuel_channels in the outer portion. The designer also might choose to pin graphite blocks to the inside of the vessel so that the reflector moves with the vessel, creating an annulus between the core and the reflector, 286 A third possibility is a combination of the above two methods, creating an annulus between two reflector regions. Calculations were made to‘determine'the relative characteristics of the MSCR with (a) no reflector, (b) a 15-in. reflector outside the annulus, (¢) a 7.5-in, graphite region between the core and annulus and a 7.5-in. ~graphite regidn outside the annulus. The core composition was determined by an iterative procedure to achieve equilibrium with respect to a 2 ftalday fuel processing rate. “'Results The results are summarized in Table M,2. Here the materials cost is the sum of the inventory and replacement costs., The leakage includes all neutrons which escape from the system, are captured in the vessel, or are captured in the reflector, Table M.2, Materials Cost and Nuclear Characteristics of MSCR as. .a Function of Reflector Condition Materials Cost Conversion Leakage Case ~ Reflector .. - (mills/kwhr)@ Ratio 1 None 0.783 0.828 2,62 2 Outside annulus 0.7u40 0.8u49 | 1.90 3 Between core and annulus 0,805 0.812 3.16 y Between core and annulus 0.784 0.823 2,81 'aElectrical. These results show that the reflector outside of the annulus improves performance slightly, but that the other reflected reactors have poorer performance, This behavior can be clarified by examining the power density distributions shown in Fig. M.1l. o B The reactor with the reflector outside the annulus shows a large peak in power in the annulus because of the large thermal flux from the reflector. o ' o ORNL-LR-DWG. 69530 _ _ Fig. 1lc o . Unreflected -~ Case 1 Reflector between core and annulis -~ Case 3 1. Opmeemey— — . 1.0 - | 1N - | .8 '8._ T 1 N T N p/e | | | P/P 12 ey \\ -2 ~ ol L1 1 1 1 1 N 0 0 40 80 ' 120 ‘160‘ 2OO N24O 280 320 360 O 40 80 120 160 200 240 280 320 360 e ?. ‘jRadiuS'(in.)' Radius (in.) . . ‘ . xR | Fig. 1b - Fig. 1d 1.0y ‘Reflector outside annulus — Case 2 OReflector on both sides of annulus — Case 4 .8 : ‘ _ ' -8 .6 \\ .6 \\ P/P | | P/P_ \ ° .4 \\\\\ -4 \\\\\ ‘.2 \ .2 s ob—L 0 40 O 40 80 120 160 200 240 280 320 360 Radius (in.) Fig. M.l. 80 120 160 200 240 280 320 360 Radius (in.) Power Density Distribution in MSCR. 288 Since a peak in the fission rate occurs, the source of neutrons aimed out of the reactor is increased; however, the reflector returns many of the neutrons leaving the annulus. The net effect, as shown in Table M.2, is a slight reduction in leakage. H The reactor with the reflector between the core and annulus experi- ences a considerable flattening of the power distribution. The fission rate in the annulus remains large and furnishes a large source of neutrons adjacent to the reactor periphery. This source is larger than for the un- reflected case, and a higher leakage results, | The reactor with the graphite regions on each side of the annulus com- bines the bad features of cases 2 and 3. The fission rate is large in the annulus, giving a large source for neutrons to the reactor periphery, Thus it is seen that theimain reason that a reflector has such a small effect is the presence of the fuel annulus., Any addition of reflec- tor increases the fission rate in this region. These fission neutrons are close to the reactor periphery, where they may be absorbed in the vessel or leak out of the reactor, A reflector would be much more beneficial if a design which eliminated the fuel annulus could be devised. Conclusions Use of a reflector in the MSCR improved the reactor performance only slightly [0.02 increase in conversion ratio and 0.04 mills/kwhr (electrical) decrease in fuel cycle costs]. Of the reflected-reactor configurations considered, the reactor with the reflector outside of the fuel_ahnulus was the only one which improved performance., However, a method of pinning graphite to the vessel without leaving large cracks is unknown. Also, the extra cost of fabricating an INOR-8 vessel 2.5 ft larger is unknown. Therefore, in view of the slight benefit to be gained and the added design uncertainties and complexity, it appears that no reflector should be used in the MSCR study. If these uncertainties should be removed and slight improvements in performance become important, the gains available with a reflector should be exploited. in ) a L3 289 Appendix'N DETAILED ESTIMATE OF 1000 Mwe MSCR CAPITAL INVESTMENT™ C. H, Hatstat** Summarz The cycle chosen for this analysis is shown in elementary form in Fig. 4.3. A 2500 Mwt reactor is cooled with a fuel-bearing molten salt, from which heat is transferred to an inert salt in eight vertical shell- and-tube heat exchangers; the heat in the inert salt is removed in a sys- tem of 16 shell-and-tube superheaters and eight reheaters, of a design similar to that of the superheaters. Approximately 63% of the superheated steam flows to a system of four Loeffler boilers, where it produces satu- rated steam by mixing with the tufbine feed water. The remaining super- heated steam is delivered at 2400 psi, 1000°F; to the throttle of the steam turbine, Exhaust steam from the high-pressure turbine elements. is.reheated to 1000°F in the reheaters and flows to the intermediate pressure turbine, from which it flows to a 6-flow low-pressure unit. Condensate is returned to the Loeffler boilers through eight stages of feed-water heating., The _gross power output of the cycle is 1083 Mwe at a throttle steam flow of approximately 8 x 105 1b/hr and a condenser pressure of 1.5-in. Hg. - The plant design is based on an Atomic Energy Commission reference Site in Western Massaehusetts.';The site is assumed to have an,adequate source of :circulating wa_ter'for_':the t_urbine. Because of the low vapor preéSure of the reactor'cooiant, nigh-pressure containment is not consid- ered necessary; the reactor and its auxiliaries are contained in a sealed, 7 steel lined concrete structure which forms a part of a subd1v1ded biolo- , glcal shield with a total thlckness of 10 feet. The turblne-generator and the other components of the steam—condensate 'system are housed 1n a conventlonal steel frame bulldlng,r The turbine *Extracted from SL-1554, SL-19%4. *¥Sargent and Lundy, Engineers, Chicago, Illinois. 290 building and the reactor building are arranged so that one traveling bridge crane services both buildings.‘ Other structures on the site which are included in the cost estimate are the crib house, circulating water intake and discharge flumes and tunnels, waste disposal building and stack, and foundations for oil and condensate tanks. Road and rail access are also pro#ided for the plant. The cost estimate includes all systems and components necessary for a complete plant, In addition to the energy conversion components, the following equipment and/or systems are estimated in detail, 1, Radioactive waste treatment and disposal systems and building. 2, Cover gas supply and distribution system. - . 3. Reagent gas supply and-disposal system, 4, UFu addition facility., 5. Fuel salt handling, sampling and storage systems. 6. Reactor vessel and primary pumps. | 7. Thermal shield and cooling system, 8. Emergency shutdown cooling system. 9, Reactor control system, 10, Fuel salt chemical treatment system, 11, Intermediate salt chemical treatment system. 12, Intermediate salt handling, sampling and storage systems, 13. Coolant pump lubricating oil systems. 14, Hot sampling facilities. 15, Remote maintenance facility., _ , 16, Subdivided shielded areas for reactor auxiliaries in the reactor building. Investment Requirements The capital investment required for the concept which is described in this report has been estimated on the basis of preliminary design and material quantities prepared by Oak Ridge National Laboratory and Sargent & Lundy. The estimating data for the heat cycle, auxiliary systems, and o’ primary and intermediate system components were prepared by Oak Ridge I ) 291 National Laboratory. The cost estimaté was prepared by Sargent & Lundy, using accounting pfocedures specified by the U, S, Atomic Energy Commission in the Guide to Nuclear Power Cost Evaluation. The direct construction cost and indirect cost are summarized below. The detailed estimate is presented on subsequent pages. Estimated Direct Construétion Cost $ 89,341,200 Indirect Costs 48,582,600 Total Capital Investment for $137,923,800 Structures and Equipment Coolant Salt Inventory plus 10,951,800 1.5% for Interest During Construction Total Investment, Excluding $148,875,600 Fuel Salt Table N.1. 1000 Mwe Molten Salt Converter Reactor Plant — Estimate of Capital Investment ONE (1) 2500 MWt MOLTEN SALT REACTOR ONE (1) 1000 MWe REHEAT TURBINE GENERATOR UNIT C.C.6F 40" L.S.B. (2400 Psi. - 1000°F - 1000°F) (Prices as of 11-1-62 and Based on a 40 Hour Work Week) QUANTITY MATERIAL OR LABOR EQUIPMENT ACCOUNT 21 - STRUCTURES & IMPROVEMENTS 211 Ground Improvements .1 Access Roads for Permanen Use - .11 Grading ) .12 Surfacing ) .13 Culverts ) .14 Bridges & Trestles ) 15 Miles .15 ~ Guards & Signs ) .16 Lighting W2 General Yard Improvements .21 Grading & Landscaping Lot $6,000 $19,200 .22 Roads Sidewalks & Parking : , Areas 47,000 SF 16,500 7,600 .23 Retaining Walls, Fences & Railings - 231 Fence, Post, Gates 2,450 LF 8,500 3,200 TOTALS cé¢c -In Place $25,200 24,100 11,700 O Table N.1 {continued) QUANTITY MATERIAL OR _ T ' EQUIPMENT ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) 211 Ground Improvements (Cont'd,) 2 General Yard . ' - Improvements (Cont'd.) .24 Outside Water Distribution .~ Systems Including Fire - Hydrants & Water Tanks for .. Genéral Use = = " +241 Domestic Water System 2411 500 G,.P.M. Deep Wells,) - " Including Pump & o Accessories .2412 Storage Tank, 300 Gal. & Controls .2413 Water Softener, ' Piping & Controls .2414 Piping Lot 13,000 S N Nt St N N N .242 TFire Protection System .242]1 Water Storage Tank .2422 2000 GPM Fire Pump & Motor Drive .2423 Other Fire Protection : Equipment 2424 Piping, Including Hydrants .2425 Hose & Hose Houses Lot 27,500 S’ Nt S N Nt Nt S Nt LABOR 17,600 27,500 TOTALS 30, 600 55,000 £6¢ ) Table N.1 (continued) QUANTITY ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) Ground Improvements (Cont'd.) 211 .2 .25 .251 .252 .2521 .2522 .2523 .2524 253 .2531 .2532 .2533 .2534 .2535 .26 .261 .262 .263 .264 General Yard Improvements (Cont'd.) Sewers & Drainage Systems: Yard Drainage & Culverts Sanitary Sewer System Septic Tank Dosing Syphon ) Distribution Box ) Tile Field (Drainage) ) Storm Sewer System: Excavation & Backfill ) Vitrified Clay Tile ) (6" & 8") . ) Reinforced Concrete Pipe ) (27" & 30") ) Manholes ) Outfall Structure ) Roadway & General Lighting Security Fence Lighting) Roadway Lighting ) Parkway Cable ) Trenching for Parkway ) Cable ) Lot Lot Lot Lot MATERIAL OR EQUIPMENT 4,000 $12,000 13,000 8,000 LABOR 7,000 18,400 11,200 11,200 TOTALS 11,000 30,400 24,200 19,200 - %62 Table N.1 {continued) QUANTITY MATERJAL OR : ' | EQUIPMENT ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) 211 Ground Improvements (Cont'd.) .3 Railroads =~ = 311 Grading | ) .312 Bridges, Culverts & Trestles) 5 Miles 135,000 .313 Ballast & Track ) .314 Signals & Interlocks ) +32 On Site .321 Ballast & Track 265 LF 1,500 © TOTAL ACCOUNT 211 §245,000 212 Buildings =~ . '~ 212A Turbine Generator Building '~ Including Office, Control Room, Cable Room, Switch Gear Room .1 Excavation & Backfill .11 = Earth Excavation 11,500 CY - .12 . Rock Excavation 5,650 CY - .13 . Backfill : 6,350 CY 2,000 .14 Disposal R 10,800 CY - .15 Dewatering . Lot - .3 . Substructure Concrete .31 Forms ) .32 Reinforcing ) .33 Concrete ) .34 Waterproofing ) 6,750 CY 232,000 .35 Patch & Finish ) Conc. .36 Miscellaneous Anchor Bolts,) Sleeves Etc. Embedded in ; ‘Concrete LABOR 132,000 1,600 $256, 500 11,500 45,200 9,400 4,400 75,000 218,400 TOTALS 267,000 3,100 $501,500 11,500 45,200 11,400 4,400 75,000 450,400 Gg6c Table N.1 (continued) QUANTITY MATERIAL OR LABOR TOTALS EQUIPMENT ACCOUNT 21 ~ STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212A Turbine Generator Building Including Office, Control Room, Cable Room, Switch Gear Room (Cont'd.) ' Superstructure 41 Superstructure Concrete 411 Forms ) ' , 412 Reinforcing ) 34,000 SF $57,500 $49,000 $106,500 413 Concrete ) of Floor ‘ 42 Structural Steel & Miscellaneous Metal 421 Structural Steel 1,650 T 535,000 128,000 663,000 422 Stairs, Ladders, ' Railings, Walkways, Gratinga, Etc. Lot 55,000 24,000 79,000 .43 Exterior Walls 431 Masonry - : - - 432 Insulated Metal Siding 66,400 SF 134,000 46,400 180,400 44 Roofing & Flashing 441 Pre-Cast Roof Slabs ) 442 Built-Up Roofing & ) Flashing ) .443 Poured Concrete Roof ) 35,600 SF 32,000 ‘ 36,000 68,000 Deck ) | ‘ 444 Insulation ) O 92 o s R P Ao S 1 P B LSS ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) Table N.1 (continued) QUANTITY MATERTAL OR EQUIPMENT 212 Buildings (Cont'd.) | 212A Turbine Generator Building Including Office, Control Rocm, Cable Room, Switch Gear Room A 045 451 462 ‘047_" 471 472 473 474 .48 481 482 49 491 .492 .5 Glazed Tile Plastering Including ‘Superstructure (Cont'd.) Interior Masonry & | Partitions : ~ Structural Tile 29,800 ST $15,100 .46 Doors & Windows‘ A6 Doors = . Lot 11,500 Windows: IR 12,600 SF 48,000 Wall and Ceiling ' Finish Metal Ceiling | 6,200 SF 5,000 Lathing and Furring Acoustical Tile Floor Finish e N N’ et Sl ‘Cement. ) Tile ) Lot 30,000 Painting Glazing and Insulation Painting Lot 10,500 Glass and Glazing - - Stack (Heating Boiler and Auxiliary Boiler ) 1 4,000 M LABOR $20,300 4,400 20,000 4,400 38,100 32,400 1,600 TOTAL $35,400 15,900 68,000 9,400 68,100 42,900 Incl, 462 5,600 " 62 ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212A Turbine Generator Building Including Office Control Room, Cable Room, Switch Gear Room (Cont'd.) .6 .61 611 612 .613 614 .615 .62 621 .622 .623 624 .625 .626 627 .628 6221 .6222 .63 .64 .641 .642 .643 Building Services Table N.1 QUANTITY Plumbing & Drainage Systems ) Plumbing ) Drainage ) Lot Duplex Sump Pump ) Domestic Cold Water Tank ) Domestic Hot Water Tank ) Heating Boiler & Accessories Heating Boiler ) Unit Heaters ) Discharge Ducts Condensate Pump & Receiver ) Flash Tank ) Lot Piping ) Fuel 0il Transfer Pump ) Heating 011 Tanks - Day ) & Storage ) Berm for Fuel Oil Storage ) Tank ) Foundation for Heating 0il ) Day Tank ) Ventilating System Air-Conditioning System ) Air-Conditioning Control Room ) ; . Office Air-Conditioning ; Laboratory Air Conditioning (continued) MATERIAL OR EQUIPMENT $60,000 77,000 55,000 LABOR $32,000 50,400 28,000 e TOTALS $92,000 .N O o0 127,400 83,000 » . i * 1) . P Table N.1 (continued) QUANTITY MATERIAL OR - | | EQUIPMENT ACCOUNT 2] - STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212A Turbine Generator Building Including Office, Control Room, Cable Room, Switch Gear Room (Cont'd.) .6 Building Services (Cont’'d.) .66 Lighting & Service Wiring . .661 Control Panels & Cabinets .662 Conduit . - .663 Wiring =~ - .664 Fixtures Switches & Receptacles .67 Fire Protection System (Water ' ~ Lines, Hose, Sprinkler, Etc.) Lot 12,000 TOTAL ACCOUNT 2124 §1,422,100 ) ; Lot $46, 500 ) 212D Waste Disposal Building .1 Excavation and Backfill .11 Excavation .111 Earth 85 c.y. - .112 Rock _ - - .12 Backfill ‘ .121 Earth — ' 45 c.y. - .13 Disposal _ , ‘ .131 Earth ) | .50 c.y. - ".132‘ Rock' ) - <15 Dewatering : »151 Pumping Lot - .3 Substructure Concrete ‘ Including, Forms, Anchored Steel .31 Bottom Slab ) .32 Walls to Grade ) 100 c.y. 3,500 LABOR $39,200 2,400 $920, 500 100 100 50 1,000 3,200 TOTALS $§85,700 14,400 et a Y $2,342,600 100 100 50 1,000 6,700 a) 662 Table N.1 (continued) QUANTITY ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) O 212D Waste Disposal Building (Cont'd.) A Superstructure .42 Structural Steel and Miscellaneous Steel 421 Structural Steel and Girts 105 Tons 422 Miscellaneous Steel Galleries Stairs, landing, Handrailing, Ladders, Etc. Lot .43 Exterior Walls .431 1Insulated Metal Siding 6,800 s.f. .44 Floors, Barriere, Including Reinforcing, Forms, Etc. 441 Walls Above Grade 135 c.y. 442 Floors 80 c.y. 4421 Pre-Cast Roof Slab 2,200 s.f. .45 Interior Masonry and Partitions 46 Doors and Windows .461 Doors , Lot 462 Windows 1,200 s.f. .48 PFloor Finish (Cement) 3,500 s.f. .49 Exterior and Interior Finishes 491 Painting Floor and Walls 492 Painting Structural and Miscellaneous Steel .493 Heavy Duty Coating ) Lot 494 Exterior Coating Below Grade ; MATERIAL OR EQUIFMENT $34,000 4., 500 13,600 6,500 3,200 2,000 2,000 4,500 500 25,000 LABOR $8,000 2,000 4,900 4,000 1,800 2,000 1,000 2,000 1,500 20,000 TOTALS $42,000 6,500 18,500 00¢ 10,500 5,000 4,000 3,000 6,500 2,000 45,000 ¥ * Table N.1 {continued) QUANTITY ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) - 212D Waste Disposal Building (Cont'd.) " .6 - Building Services .61 . Plumbing and Drainage . 'System. Lot .66 Lighting and Service : Conduit - . Lot .67 Fire- Protection System Lot .~ TOTAL ACCOUNT 212D 212F Miscellaneous Structures .1 Gate House Lot .2 Electrical Lot .3 Waste Storage Pond Each TOTAL ACCOUNT 212F 212G Reactor Plant Building : .1 - Excavation & Backfill .11 ~ Earth Excavation 5,655 c.y. .12 Rock Excavation 1,090 c.y. .13 Backfill 755 c.y. .14 Disposal 5,990 c.y. .15 Dewatering Lot 3 MATERIAL OR EQUIPMENT $4,800 1,200 4,000 $109,300 $5,500 3,000 2,800 $11,300 LABOR $2,200 1,800 1,000 $56,650 $5,200 2,800 5,200 $13,200 $5,700 21,800 1,200 2,400 55,000 TOTALS $7,000 3,000 5,000 $165,950 $10,700 5,800 8,000 $24,500 $5,700 21,800 1,200 2,400 55,000 10 ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212G Reactor Plant Building (Cont'd.) .3 .31 .32 .33 .34 .35 .36 Substructure Concrete Forms Reinforcing Concrete Waterproofing Patch & Finish Miscellaneous Anchor Bolts, Sleeves Etc. Embedded in Concrete Superstructure Superstructure Concrete Forms ) Reinforcing ) Concrete Interior ) Structural Steel & Miscellaneous Steel Structural Steel & Reactor Supports Stairs, Ladders, Railings, Walkways, Grating, Etc. Exterior Walls Masonry Insulated Metal Siding Concrete Walls Teble N.1 (continued) T S N N’ N N Nt Nt QUANTITY 5,730 c.y. 6,981 c.y. 964 T Lot 54,100 s.f. 5,150 c.y. MATERIAL OR EQUIPMENT $200,000 365,000 305,000 30,000 110,000 250,000 LABOR $185,000 225,000 80,000 14,500 38,000 157,000 TOTALS $385,000 590,000 385,000 44,500 148,000 407,000 co¢ - 212 212G Table ACCOUNT 21 - STRUCTURES & IMPROVEMENTS (Cont'd.) Buildings (Cont'd.) Reactor Plant Building (Cont'd.) .4 42 443 45 451 453 46 461 462 48 .481 .49 491 492 493 S5 .6 i61 611 .612 613 L4 Superstructure (Cont'd.) ‘Roofing & Flashing Pre-Cast Roof Slabs ) Built-Up Roofing & Flashing) .\ Insulation ) - Interior Masonry & Partitions Structural Tile Hot Cells | Doors & Windows Doors- ‘ Windows ‘Floor Finish ‘Cement Painting Glazing Insulation Painting Glass and Glazing Insulation of Reactor Chamber ‘Stack (When Supported on Building) 'Building Services Plumbing & Drainage System Plumbing ) Drainage ) Sump Pump ) N.1l QUANTITY 23,300 s.f. Lot Lot 7,250 s.f. 35,000 s.f. Lot Lot (continued) MATERIAL OR LABOR EQUIPMENT $21,000 $21,600 400,000 80,000 2,500 1,200 27,000 12,000 12,000 14,400 7,000 18, 500 Included .462 Incl, in Acct, 221.32 Incl., in Acct. 212A 15,000 8,000 TOTALS $42,600 480,000 3,700 39,000 26,400 25,500 23,000 e0¢ Table N.1 (continued) QUANTITY ACCOUNT 21 - STRUCTURES & JMPROVEMENTS (Cont'd.) 212 Buildings (Cont'd.) 212G Reactor Plant Building (Cont'd.) .6 .62 .63 .66 .661 .662 .663 .664 .67 Buildings (Cont'd.) Cooling System ) Ventilating System ) Lighting & Service Control Panels & ) Cabinet ) Conduit ) Wiring ) Fixtures, Switches ) & Receptacles ) Fire Protection System (Water Lines, Hose, Sprinkler, Etc.) TOTAL ACCOUNT 212G TOTAL ACCOUNT 212 218 Stacks 218A Concrete Chimney .1 .2 4 6 O Excavation and Backfill Substructure Concrete Concrete Chimney Obstruction Lighting TOTAL ACCOUNT 218A TOTAL ACCOUNT 218 TOTAL ACCOUNT 21 Lot Lot Lot » MATERIAL OR EQUIPMENT $130,000 17,000 8,500 '$1,900,000 $3,442,700 15,000 $15,000 $15,000 $3,702,700 LABOR TOTALS $70,000 $200, 000 19,600 36,600 1,500 10,000 $1,032,400 $2,932,400 $2,022,750 $5,465,450 16,000 31,000 316,000 $31,000 $16,000 $31,000 $2,295,250 $5,997,950 70 Table N.1 QUANTITY ACCOUNT 22 - REACTOR PLANT EQUIPMENT 221 Reactor Equipment : .1 Reactor Vessel and Supports .11 'Reactor Vessel Supports Lot .12 Vessel and Internals ) .13 Pump Suction Columns ) Lot .14 Graphite Rods ) .15 Heaters - -+16 Insulation .2 Reactor Controls = .21 Reactor Control Salt Addition : ‘Tank = 1 +22° Reactivity Control Drain Tanks 2 +23 Drain Tank Condenser 1 .231 Condensate Pump 2 .24 = BF3 Injection System .241 BF3 Cylinders - .25 Piping, Valves, Etc. .3 Reactor Shielding .31 Thermal Shield System .311 Thermal Shield & Supports 1 .312 Surge Tank, 2000 Gal. 1 .313 Circulating Pumps 2 " .314 Heat Exchanger 1 - 315 Viping, Valves, and Insulation Lot 1 ~ +32 Biological Shielding - " " Insulation, Shield Plugs, Etc. o Lot .34 Shield Cooling System .341 Closed Loop Liquid System ~ .3411 Shield Cooling Heat Exchanger (4000 Ft. 2 Surface - Admiralty) 2 (continued) MATERIAL OR EQUIPMENT $15,000 7,540,000 37,500 7,000 6,200 42,400 1,200 300 LABOR $8,000 560,000 Included 6,400 100 800 100 100 Not Included Incl., Account 228 75,000 2,400 2,600 25,000 16,000 200 200 500 Incl. Account 228 255,000 35,000 124,000 2,500 TOTALS $23,000 8,100,000 37,500 13,400 6,300 43,200 1,300 400 91,000 2,600 2,800 25,500 379,000 37,500 G0¢ Table N.1 (continuved) QUANTITY MATERIAL OR LABOR TOTALS EQUIPMENT ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) 221 Reactor Equipment (Cont'd.) Reactor Shields (Cont'd.) 34 Shield Cooling System (Cont'd.) .341 Closed Loop Liquid System (Cont'd.) .3412 Shield Cooling Circulating Pumps & Motors (2500 GPM 75 HP Motor) 3 $7,500 800 8,300 .3413 Piping & Valves Lot Included in Account 228 .3414 Cooling Coils Embedded in Concrete (16000 Ft. 1" Steel) Lot 17,500 32,500 50,000 .3415 H20 Storage Tank 3000 Gallons 1 1,200 300 1,500 .7 Reactor Plant Cranes & Hoists Included in Account 251 TOTAL ACCOUNT 221 $8,070,800 - $752,500 $8,823,300 222 Heat Transfer Systems .1 Reactor Coolant Systems .11 Reactor Salt Circulating Pumps 9075 GPM Including 1600 HP Motors 8 2,768,000 25,000 2,793,000 .12 Reactor Salt Piping Lot 215,000 20,000 235,000 .13 Insulation Lot 5,000 5,600 10,600 .2 Intermediate Coolant System : .21 Pumps Including Supports -+211 Coolant Salt Pumps - 13,900 GPM Including 8 3,640,000 30,000 3,670,000 ) 2,000 HP Motors : : : .212 Auxiliary Pumps and Drives - - - . - +213 Insulation Included in Account 228 90¢ Table N.1 (continued) QUANTITY ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) 222 Heat Transfer Systems (Cont'd.) .2 Intermediate Coolant System (Cont'd.) .22 - Intermediate Coolant Piping _ and Valves .221 Pipe, Valves, Supports, Etc. Lot .222 Insulation. = : Lot .223 Coolant Salt Drain Tanks - Including Heaters and Insulation -2 .23 Primary Heat Exchangers - & Supports 8 .3 Steam Generators Superheaters & Reheaters. o «31 Loeffler Boilers 4 .32 Superheaters - | 16 .322 Steam Reheaters 8 .35 Auxiliary Start-up Boiler (300 Psi. 50,000 1b/nhr. oil Fired) 1 .36 Insulation for Above Equipment A Reactor Coolant Receiving Supply and Treatment .411 Fertile Salt Addition System .4111 Fertile Salt Addition Tank 1 .42 Reagent Gas System 421 H, Supply .422 HF Supply 423 Piping 431 Reactor Salt Purification System .4311 Chemical Treatment Tank 1 4312 Fertile Salt Storage Tank 1 MATERJAL OR EQUIPMENT $1,925,000 50,000 70,000 2,320,000 4,000,000 6,350,000 1,400,000 60,000 Included in Account 228 1,200 46,600 27,500 LABOR 172,000 48,000 12,800 28,000 160,000 48,000 10,000 5,000 50 Not Included Not Included Included in Account 228 400 300 TOTALS $2,097,000 98,000 82,800 2,348,000 4,160,000 6,398,000 1,410,000 65,000 1,250 47,000 27,800 L0e Table N.1 (continued) ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) 222 Heat Transfer Systems (Cont'd.) 431 4313 432 4321 bk 441 L4411 Reactor Coolant Receiving Supply and Treatment (Cont'd.) Reactor Salt Purification System (Cont'd.,) Radioactive Salt Sampler Intermediate Salt Purification System Chemical Treatment Tank Reactor Salt Charge System Reactor Salt Preparation System Salt Melt Tank and Appurtenances 4412 UF, Addition System 4h2 4421 Intermediate Salt Charge System Intermediate Salt Preparation Tank and Appurtenances QUANTITY Lot MATERIAL OR EQUIPMENT $67,500 19,200 30,000 3,000 30,000 LABOR $1,600 200 1,000 500 1,000 TOTALS 80¢ $69,100 19,400 31,000 3,500 31,000 Table N.1 {(continued) QUANTITY ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) 222 Heat Transfer Systems (Cont'd.) 4 Reactor Coolant Receiving . " Supply and Treatment (Cont'd.) .45 Cover Gas Supply and Purification System 4531 Dryer: | _ 4532 Heater ' - 4533 02 Removal Unit 4534 Coolers 4535 Pure Hg Reservoir 4536 09 Analyzer 454 Piping .5 Intermediate Coolant Storage Tanks, Etc. 2 TOTAL ACCOUNT 222 =N 223 Fuel Handling and Storage ' Equipment .31 Fuel Salt Drain and Storage System * .311 Drain Tanks & Cooling Jacket 54 .312 Drain Tank Condenser 1 23121 Condensate Pump 2 .313 Decay Storage Tank Including : Cooling Jacket 5 .314 Fuel Withdrawal Transfer Tank 1 MATERIAL OR EQUIPMENT - 200 400 3,000 200 400 7,000 Included Acct. 228 70,000 $23,039,200 $1,269,000 8,000 1,400 55,000 37,000 LABOR TOTALS 50 250 100 500 200 3,200 100 300 100 500 500 7,500 2,000 72,000 $570,500 $23,609,700 $16,800 $1,285,800 300 8,300 100 1,500 1,000 56,000 100 3,800 60€ 223 225 ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) Fuel Handling and Storage Equipment Table N.1 (continued) QUANTITY +321 Puel Withdrawal Metering Tanks 1 .33 Flush Salt Storage Tanks .34 Piping TOTAL ACCOUNT 223 Radioactive Waste Treatment & Disposal .1 Liquid Waste Systems .11 High Level Storage Tank .111 H.L. Storage Tank Pump .112 H.L. Waste Evaporator .113 H.L. Waste Condenser .114 Evaporator Recycle Pump .13 Demister .14 H.L. Concentrated Waste Storage Tank .15 KOH Scrubber .151 KOH Make-up Tank & Pump .16 H, Burner .161 H, Burner Condenser .17 P%ping .2 Gas Waste .21 Stack Blower .211 Absolute Filter, - .212 Roughing Filter o B et = B o N 1,200 200 MATERIAL OR LABOR TOTALS EQUIPMENT ' $6,200 $300 $6,500 152,500 2,800 155,300 Included in Account 228 $1,495,800 $21,400 $1,517,200 12,400 600 13,000 400 100 500 2,000 100 2,100 - 600 50 650 400 100 500 600 ) 50 650 2,500 150 2,650 12,100 400 12,500 1,400 100 1,500 225 75 300 1,000 100 1,100 Included in Account 228 ' 20,000 2,400 22,400 4,000 400 4,400 1,400 Ote ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) 225 Radiocactive Waste Treatment & Disposal (Cont'd.) .22 221 .222 - .223. .23 .231 232 233 234 W24 241 .242 243 .244 «25 .26 .27 .28 .281 T;29 H.F. Absorbers Including Charcoal L Absorber Coolers Vacuum Pump = H.F. Absorber Vacuum Tank Air Cooled Absorbers Including Charcoal - 1.5" Finned Tubes 3" Finned Tubes 6" Finned Tubes Dilution Air Duct & Dampers Water Cooled Absorbers 1/2" Tubes ' 1" Tubes 1-1/2" Tubes 2" Tubes Decay Tank Absorber Cooling Water Condenser Helium Recycle Compressor "BF3 Stripper Vacuum. Pump Piping =~ TOTAL ACCOUNT 225 226 Instrumentation and Controls .1 .2 Reactor Heat Transfer Systems Table N.1 (continued) QUANTITY - P NN N - = N N *} MATERIAL OR LABOR . TOTALS EQUIPMENT $16,000 $2,400 $18,400 100 _ 50 150 500 50 550 300 : 50 350 3,600 400 4,000 8,000 600 8,600 15,000 800 15,800 8,000 3,290 11,200 23,200 800 24,000 60,400 1,900 62,300 51,400 1,900 53,300 49,600 1,800 51,400 12,500 800 13,300 900 - 100 1,000 2,000 200 2,200 28,000 2,400 30,400 500 50 550 Included in Account- 228 $338,825 $22,325 $361,150 $300,000 $170,000 $470,000 70,000 50,000 120,000 e ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) Table N.1 (continued) QUANTITY 226 Instrumentation and Controls (Cont'd.) 2217 .3 4 Service to Fuel Handling and Storage Service to Radioactive Waste & Disposal Radiation Monitoring Steam Generators Control & Instrument Piping & Wiring Electrical Connections Other Miscellaneous TOTAL ACCOUNT 226 Feed Water Supply and Treatment .1 .2 .21 22 .23 24 Raw Water Supply Make-up Water Treatment Evaporator Ion Exchange Equipment, Filters, Etc. Acid & Caustic Transf. Pumps & Drives Demineralized Water Storage Tanks Caustic Tank Acid Tank ' Foundation Piping & Valves Insulation Steam Generator Feed- Water Purification 1 Lot 1 Lot MATERIAL OR LABOR TOTALS EQUIPMENT . $120,000 $80,000 $200,000 60,000 50,000 110,000 120,000 80,000 200,000 Included in Account 235 Included in Account 235 ‘Included in Account 235 Included in ‘Account 235 $670,000 $430,000 $1,100,000 Included in Account 211 $45,000 $10,000 $55,000 600 200 800 30,000 Included 30,000 2,200 400 2,600 2,200 400 2,600 3,500 2,500 6,000 . Included in Account 228 Included in Account 228 O cle Teble N.1 (continued) ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) Feed Water Supply and ‘Treatment (Cont d.) 227 WAL b2 421 422 w423 424 425 427 S .51 «3511 .512 .52 .521 .522 .. 53 .531 Feed-Water Heaters Deaerating Heaters - "E" 4,020,000 #/Hr, 150 Psig. CIOSed Heaters L.P. Heater "A" . L.P.=Heater‘"B" L.P. Heater "C" L.P. Heater "D" -H.P. Heater "F" H.P. Heater "G" H.P. Heater "H" . Feed-Water Pumps and Drives Feed-Water Pumps & Drives 6600 GPM Pumps - 2465 Psig Hd. 11,300 H.P. - B.F. Pump Turbine Drive 5600 RPM Motor Driven Start-Up F.W, Pump 6000 GPM Pump 850 Psig. Hd. 3500 H.P, Start-Up FW Pump Motor Heater "A" Drain Pumps and Drives . 620 GPM Pump 285 Psig, Hd. ) 125 H.P. Heater "A" Drain ) Pump Motor : ) QUANTITY N Wiwwwwww ) MATERIAL OR EQUIPMENT $240,000 75,000 63,000 63,000 81,000 315,000 429,000 441,000 405,000 750,000 70,000 55,000 22,500 LABOR $15,000 5,000 5,000 3,000 3,000 3,000 3,000 3,000 12,000 30,000 3,000 2,000 TOTALS $255,000 80,000 68,000 66,000 84,000 318,000 432,000 444,000 417,000 780,000 73,000 57,000 24,000 eTe ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) 227 Feed Water Supply and Treatment . 228 (Cont'd.) o3 Feed-Water Pumps and Drives (Cont'd.) .54 Heater "C" Drain Pumps and Drives .541 700 GPM Pumps 210 Psig. Hd.) .542 100 H.P., Heater "C" Drain ) Pump Motor ) 55 Boiler Steam Circulators and Drives .551 5,300,000 #/Hr. Steam Circulator - 2500 #675°F .552 5000 H.P. Turbine Drive for Steam Circulator 500 # Steam; 10,000 RPM .553 5000 H.P. Motor for Steam Circulator Including Gear and Mag. Coupling TOTAL ACCOUNT 227 Steam, Condensate, Feed Water, and all Other Piping, Valves Etc. - For Turbine Plant, Crib House and Other Reactor Plant Auxiliaries .1 .11 .12 .13 Pipe, Valves, Fittings, Etc. Turbine Plant ) Other Interior Piping ) Yard Pipe Etc. ) Teble N.1 (continued) QUANTITY 1 Lot MATERIAL OR EQUIPMENT $30,000 975,000 560,000 __ 100,000 $4,758,000 $4,025,000 LABOR TOTALS $2,500 $32,500 40,000 1,015,000 30,000 590,000 7,000 107,000 $181,500 $4,939,500 $2,575,000 $6,600,000 4’2 ACCOUNT 22 - REACTOR PLANT EQUIPMENT (Cont'd.) Steam, Condensate, Feed Water, and all Other Piping, Valves Etc. 228 229 - = For Turbine Plant, Crib House and Other Reactor Plant Auxiliaries (Cont'd.) 2. Insulation L .21 Piping Insulation .22 Equipment Insulation TOTAL ACCOUNT 228 Other Reactor Plant Equipment W2 Remote Maintenance Facilities 4 Coolant Pump Lube 0il System .41 Storage Tank ) 42 Pumps ) 43 0il Coolers ) 44 Filters ) .45 Piping .62 Intermediate Salt Sampling ~ System ' ‘ .621 Sampler and Appurtenances TOTAL ACCOUNT 229 TOTAL ACCOUNT 22 Table N.1 (continued) QUANTITY 1 Lot 1 Lot Lot NN = Lot Lot [y MATERIAL OR EQUIPMENT $455,000 155,000 $4,635,000 3,000,000 29,400 16,500 $3,045,900 LABOR TOTALS $520,000 $975,000 195,000 350,000 $3,290,000 $7,925,000 Included 3,000,000 600 30,000 Included in Account 225 2,000 18,500 $2,600 $3,048,500 $5,270,825 $51,324,350 GTe QUANTITY ACCOUNT 23 - TURBINE GENERATOR UNITS Turbine Generators . Turbine Foundations .11 Concrete - Including Reinforeing Steel, Ete. 1 Lot «1l2 Miscellaneous 1l lot 2 Turbine Generators .21 Turbine Generator Units - As Follows: 1000 MWe Reheat Turbine Generator Unit C.C.6F. 40" L.S5.B. Complete with Accessories Steam Conditions 2,00 Pai - 1000*F =1000°F Generators: 1,280,000 KVA Total .85 P.F. and .64 SCR 1 «22 Accessories - Other Than Standard 23 Generator .24 Exciter (Motor Driven) 3 Reserve Exciter 1 TOTAL ACCOUNT 231 Circulating Water System 1 Pumping and Regulating Systems 11 Pumps, Drives & Controls 112 134,000 GFM Vertical Mixed Flow Circulating Water Pumps Head 30 ft. 3 231 232 Table N.1 (continued) MATERIAL OR FQUIFPMENT $175,000 10,000 19,815,000 §55,310.000 360,000 LABOR TOTALS $175,000 $350,000 10,000 20,000 W P N 960,000 20,775,000 Included in Account 231.21 Included in Account 231.21 Included in Account 231.21 10. 000 50,000 18,000 378,000 Table N.1 (continued) QUANTITY MATERTAL OR LABOR TOTALS EQUIPMENT ACCOUNT 23 -'TURBINE GENERATOR ; GENERATOR UNITS (Cont®d.) 232 irculatggg Water System (Cont'd. ) Pumping and Regulating Systens | (Cont'd,) .ll Pumps; Drives & Controls (Cont'd.) " «113 1250 H.P. Motor Drive for . .~ Circulating Water Pumps 6 $270,000 $£10,000 $280,000 .«12 Traveling Screems, Etc. © +121 ' Traveling Screens complete : ‘with Motors 7 122,500 8,700 131,200 .122 1200 GPM Screen Wash Pumps . 230 Ft. Discharge Head 2 5,000 1,000 6,000 & 123 100 H.P. Mbtor for Screen ~ : - Wash Pump = 2 4,500 Included 4,500 ~«124 Trash Rake Gomplete Wlth o Appurtenances 1 27 ;500 2,500 30,000 .125 Pipe & Valves 1 Lot Ircluded in Account 228 2 Circulating Water Lines 21 Supply Lines - To Condenser - ,211 Circulating Water Piping, . Valves, Fittings, Etc. 2111 Steel Circulating Water Piping, Valves Expansion n | - Joints, Fittings, Etc. 1l 1ot 155,000 70,000 225,000 22 Discharge lines - From o Condenser .221 Circulating Water Piping, ' - Valves, Fittings, Btec. Table N.1 (continued) QUANTITY . ACCOUNT 23 - TURBINE GENERATOR UNITS (Cont'd.) 232 Circulating Water System (Cont'd.) o2 Circulating Water Lines (Cont'd.) .22 Discharge Lines - From Condenasr (Cont'd.) .2211 Steel Circulating Water Piping Valves, Expansion Joints, Fittings, Etec. 1 Lot o3 Intake and Discharge Structures .31 Intake Structures 311 River Dredging & Rock Removal l Iot .312 Intake Flume .3121 Intake Flume Proper 1 Lot .3122 Floating Boom 1 .3123 Concrete Retaining Walls 2 .313 Intake Crib House «3131 Substructure 1 Lot .3132 Superstructure - .3133 Steel 35 T 3134 Electrical Work 1 lot .32 Discharge .321 Seal Well & Discharge Tunnel Lot .322 Discharge Flume Lot O MATERIAL R BEQUIPMENT 47,000 28, 500 140,000 11,000 11,000 129,500 h,500 LABOR TOTALS Included in Account 232.21 $12,000 $12,000 45,000 45,000 10,400 17,400 29,200 57,700 135,000 275,000 4,000 15,000 13,60C 21,600 28,400 57,900 22,400 26,900 O 81¢c Table N.1 {continued) _ QUANTITY MATERIAL OR LABOR TOTALS ACCOUNT 23 ~ TURBINE GENERATOR UNITS (Cont'd.) 232 - Circulating Water System (Cont'd.) W Fouling, Corrosion Control - and Water Treatment .41 Chlorinating System . | «411 Chlorination Equipment 1 lot $45,000 $8,000 $53,000 412 Chlorine Handling Facilities 1 lot 000 2,000 000 ~TOTAL ACCOUNT 232 %*1,"224!,000" §120,200 $1,8h4,200 233 Condensers and: Auxiliaries ol Condensers: - = \ «11 Foundations - 3 47,000 $6,400 $13, 400 O +12 Condenser Shell and _ Appurtenances = .121 225,000 Sq. Ft. Single Pass : Condensers Complete with Appurtenances Including Shell, Water Boxes, Tube Sheets, Tube Supports, Hot Well, Extended Neck with Expansion Joint, Ete. 3 1,320,000 440,000 1,760,000 .13 50 Ft. long Admiralty Condenser 3 Sets 1,053,000 Included 1,053,000 Tubes \ o . .17 Instruments & Accessories 3 Sets 15,000 Included 15,000 W2 Condensate Pumps .2l Pumos & Drives .211 1875 3PM Condensate Pumps Complete with Appurtenances, Discharge Head - 325 Ft. 6 87,000 6,000 93,000 Table N.1 (continued) ACCOUNT 23 ~ TURELNE GENERATOR UNITS (Cont'd.) O 233 234 235 Condensers and Auxiliaries (Cont'd.) o2 Condensate Pumps (Cont'd.) .23, Pumps & Drives (Contt'd.) 212 LOO H.P. Motors for Condensate Pumps o2 Suction Piping o3 Air Removal Equipment and Piping «31 Steam Jet Air Ejector, with Inter & After Condensers «32 Air Suction Piping .33 Priming Bjectors TOTAL ACCOUNT 233 Central Lubricating System ol Treating & Pumping Equipment o2 Storage Tanks & Appurtensances o3 Fire Protectlon TOTAL ACCOUNT 234 Turbine Plant Boards instruments & Controls 1 Control Equipment .11 Mechanical Control Boards .12 Isolated Controller, Transmitters Ztc. 2 Isolated Recording Gauges Meters & Instruments Nt Nt s N g QUANTITY 1ot 3 888 1 Iot MATERIAL CR EQUIFMENT 846,800 100,000 10, 500 2,639,300 17,000 14,000 §31,000 $275,000 LABOR TOTALS oL, 200 $51,000 Included in Account 233.121 g, OOO 109,000 Included in Account 228 Included 10, 500 §1,65,600 $3,104,900 2,000 19,000 3,000 17,000 Included in Account 23 §5,000 36,000 $25,000 $300, 000 0z¢< o mr— i - g e A LBl 4R TGk 00 LT 1 T Table N.1 (continued) | _ QUANTITY MATERIAL OR LABOR TOTALS ACCOUNT 23 -~ TURBINE GENERATOR UNITS (Cont'd.) 235 Turbine Plant Boards Inst.nunent.s & Controls ‘ .3 . Control & Instrmnmt - - Piping & Tubing = 1 Lot $20,000 555,200 -$75,2oo of4 - Electrical Connections 1 Lot 18,000 33,600 51,600 ol TOI'AI. ACCOUNT 235 5313,000 ’ 113,600 426,600 236 Turbine Plant. Piping . : "Maln Steam Between Stop | ‘ Valves and Turbine Inlet Included in Account 231.2 .2 'Drip, Drain and Vent Piping and Valves Included in Account 228 TOTAL 236 o Included in Account 228 237 Awd.liary Equ_ment for ' Generators - .1 Excitation Panels, Switches & Rheostats Included in Account 231.2 +2 Generator Cooling Water Systems - W21 Lubncaifl.ng 0il Cooling ) . System ) .22 Generator Hydrogen ) Lot 60,000 12,000 72,000 ' Cooling System ) ' - .23 Generator Liguid ) Cooling System ) - T2¢ Table N.1 (continued) ACCOUNT 23 - TURBINE GENERATORS UNITS (Cont'd.) 237 Auxiliary Equipment for Generators (Cont'd.) .3 A 238 Other 1 .2 Central Hydrogen Cooling System Fire Extinguishing Equipment) Including Piping and COp ) System Exclusively for ) Generators ) Fire Extinguishing ) Equipment for 0il Room, Etc.) TOTAL ACCOUNT 237 Turbine Plant Equipment Gland Seal Water System Vacuum Priming System TOTAL ACCOUNT 238 TOTAL ACCOUNT 23 ACCOUNT 24 - ACCESSORY ELECTRIC EQUIPMENT 241 Switchgear .1 .11 .12 .13 14 O Generator Main and Neutral Circuits Generator Potential Transformer Compartment Surge Protection Equipment Generator Neutral Equipment Miscellaneous Items QUANTITY Lot =8N Lot MATERIAL OR EQUIPMENT $50,000 $110,000 LABOR $15,000 $27,000 TOTALS $65,000 $137,000 cce o= Included in Account 228 ~—b= wp— Included in Account 228 =i o= Included in Account 228 —t= $38,000 14,000 6,000 . 10,000 5186, 5O $4,000 1,600 800 19,200 $26,843,700 $42,000 15,600 6,800 29,200 O e At e = T Y ¢ At ok B et A i s e b R0 241 242 243 Table N.1 (continued) (Cont‘d ) 21 13 .8 KV Switchgear. - +22 4160 V, Switchgear .23 480 V, Switchgear. - TOTAL ACCOUNT 241 Switchboards : .1 Main Control Board .2 Auxiliary Power Battery & ‘ Signal Board .21 Battery & Battery Charging Panels .22 D.C. Control & Auxiliary Panels .23 A.C, Control & Instrument Panels .24 Motor Control Centers Miscellaneous Panels & Boards TOTAL Accouur 242 Protective Eguigment oL .2 General Station Grounding Equipment Fire Protection System TOTAL ACCOUNT 243 QUANT ITY ACCOUNT 24 - ACCESSORY EL ECIRIC EQUIPMENT (Cont'd.) Switchgear 2 ‘StatiOn Service Lot Lot Lot Lot b Lot Lot Lot Lot MATERIAL OR EQUIPMENT 365,000 110,000 $543,000 $82,000 15,000 18,000 7,000 80,000 16,000 $218,000 $60, 000 14,000 $74,000 " LABOR 51,200 17,600 $94,400 $31,200 5,600 4,800 1,600 13,600 11,200 $68,000 $48,800 __8,800 $57, 600 TOTALS 416,200 127,600 $637,400 $113,200 20,600 22,800 8,600 93,600 27,200 $286,000 $108,800 __22,800 $131,600 gce Table N.1 {continued) QUANTITY MATERIAL OR LABOR TOTALS EQUIPMENT ACCOUNT 24 - ACCESSORY ELECTRIC EQUIPMENT (Cont'd.) 266 Electrical Structures .1 Concrete Cable Tunnels, Compartments and Cable Trenches in Earth Lot $14,000 $21,600 $35,600 .2 Cable Trays & Supports 192,000 1b. 80,000 72,000 152,000 e3 Pipe and Steel Frames and Supports Lot 7,000 8,000 15,000 4 Foundations & Pads for Electrical Equipment Lot 5,000 5,600 10,600 TOTAL ACCOUNT 244 $106,000 $107, 200 $213,200 245 Conduit .1 Conduit .11 Power Conduit Lot $25,000 $61,200 $86,200 .12 Control and Instrument Conduit Lot 22,000 54,400 76,400 .2 Concrete Envelopes .21 10" Transite Pipe Duct Run Lot 6,000 6,800 12,800 .22 Iron Conduit Enclosed in Concrete Lot 2,000 15,200 24,200 .3 Manholes & Covers 5 5,000 5,600 10,600 TOTAL ACCOUNT 245 $67,000 $143,200 $210,200 246 Power and Control Wiring .1 Main Power Cables and Bus Duct .11 Isolated Phase Bus Duct (Generator) Lot $576,000 $49,600 $625,600 .12 Main Power Cables 1 110,000 16,000 126,000 Table N.1 (continued) QUANTITY MATERIAL OR LABOR TOTALS EQUIPMENT ACCOUNT 25 - MISCELLANEOUS POWER PLANT EQUIPMENT (Cont'd,) 251. Cranes and Hoisting Equipment (Cont'd,) ' .2 Miscellaneous Cranes and Hoists Lot $23,000 $2,000 $25,000 TOTAL ACCOUNT 251 $173,000 $22,000 $195,000 252 Compressed Air and Vacuum Cleaning ' System - L1 COmpressors and Accessories .11 . 200 C.F.M, Station Air ' Compressors Including Motor Co Drives . 2 $13,500 $1,200 $14,700 .12 250 C.F M;‘Control Air . Compressors including Motors 2 15,500 1,200 16,700 .13 Air Drying Equipment for . Control Air System 2 9,000 500 9,500 .14 = Receivers : .141 Station Air ' 2 1,300 300 | 1,600 .142 Control Air 2 1,300 300 1,600 .2 Pipe Valves and Fittings Lot «g——= Included in Account 22§ ——¥m .3 Vacuum Cleaning System Lot 16,000 4,000 20,000 TOTAL ACCOUNT 252 $56,600 $7,500 $64,100 253 .Other Power Plant Equipment - .1 Local Communication, Signal and Call System Lot $50,000 $44,800 $94,800 +2 Fire Extinguishing Equipment .21 2000 GPM Fire Pump Including _ Drive and Accessories .22 Other Fire Protection Equipment Lot $19,000 Included in Account 211.24 $1,000 $20,000 G2e Table N.1 (continued) ACCOUNT 25 - MISCELLANEOUS POWER PLANT (Cont'd.) 253 Other Power Plant Equipment (Cont'd.) o3 Furniture and Fixtures oh lockers, Shelves, and Cabtinets «d Cleaning Equipment .6 Machine Tools & Other Station Maintenance Equipment o7 Laboratory, Test & Weather Instruments «71 Radiation Monitoring Equipment 72 Miscellaneous laboratory, Test & Weather Instruments .9 Diesel Generator Unit 1000 KW Including Oil Tank TOTAL ACCOUNT 253 TOTAL ACCOUNT 25 TOTAL DIRECT CONSTRUCTION COST INDIRECT COSTS Contractorts O'HD and Profit 20% SUB-TOTAL General and Administrative 6.3% SUB-TOTAL Miscellaneous Construction 1% . SUB-TOTAL QUANTITY Lot Lot Lot lot Lot MATERIAL OR EQUIPMENT $10,000 7,000 4,000 240,000 23,000 20,000 100,000 $473,000 $78,174,625 . LABOR $10,000C 2,000 10,000 - $67,800 $97sBOQ $11,166,575 TOTALS $10,000 7,000 4,000 250,000 25,000 20,000 110,000 $540,80C $799,900 $89,341,200 2,333,300 $91,674,500 5,775,500 $97,450,000 974,500 $98,424,50C 9ce Table N.1 (continued) QUANTITY ENGINEERI NG DESIGN AND INSPECTION A - E Design and Inspect:;.on 11.1% SUB-TOTAL Nuclear Engineering 3 82 SUB-TO'I'AL L Star’c-up COs_ts o SUB—TUI'AL - Land and I.and Rights - SUB—TOTAL ‘ .Contingmcy 10% SUB-TOTAL Interest During Construction 9.4% SUB-TOTAL Fuel Charge Infémédiate Coolant Salt . Investment and 1.5% for Interest During Construction TOTAL CAPITAL INVESTMENT » MATERIAL (R EQUIPMENT LABOR TOTALS $10 945,000 $109,3h9 500 4,155,300 3113 504,800 746,900 $114,251,700 360,000 $114,611 ,700 11,461,200 $126,072,900 11,850,900 Not Included 10,790,000 161,800 $148,875,600 Leg 328 _‘APPEndix o DESIGN REQUIREMENTS FOR THE MSCR MODERATOR ‘L. G, Alexander Introduction The reactor vessel contains the moderator and holds it in a stable position during all phases of operation, provides for accepting fuel dis- charged from the heat exchangers, passage of fuel through the moderator, and discharge to the fuel pumps. The reactor must be designed to expose the fuel to neutrons at a spe- cified ratio of graphite volume to fuel volume., The graphite must be sup- ported and restrained under all circumstances, including the drained condition. Allowance must be made for differential expansion between graphite and vessel. Provision must be made fof distributing'the flow of fuel over the core entrance, and for collecting the flow at the exit. A free surface in an expansion chamber must be provided somewhere in the fuel circuit, and circulation through the expansion chamber must be maintained, Provision must be made for preheating the reactor vessel prior to charging molten salt, and for cooling the reactor vessel during operation and after shutdown., This cooling must be accomplished without the generation of ex- cessive thermal stresses. The vessel must be designed in conformance with the pressure vessel code. Means of sparging the fuel in the expansion chamber with an inert gas must be provided to remove xenon and other vola-' tile materials, Excessive thermal stress in the graphite must be avoided, and it must be composed of pieces sufficiently small that differential shrinkage due to exposure to neutrons will be tolerable in each piece, Stagnation of the fuel between édjacent blocks of graphite or between graphite and metal structure must be avoided if such stagnation leads to° excessive temperatures or stresses in either the fuel, graphite, or metal structure. Temperature at the fuel-gféphite interface should be below ‘ that at which chemical reaction, if any, takes place at an appreciable - rate, and below the temperature at which any important constituent of the Y LY [y it 329 salt (other than volatile fission products) has appreciable vapor pres- sure (e.g., if UF,, were to vaporize appreciably and diffuse into pores in the graphite, this would be disadvantageous and perhaps hazardous). Temperature gradients in the graphite near stagnation areas should not exceed those corresponding to tolerable thermal stresses., In the MSCR it is desirable, in order to reduce the graphite surface exposed to permeation by salt and fission product gases, to use moderator elements of larger diameter than in the MSRE. On the other hand, setting an allowable thermal strain of 0.001 imposes an upper limit. A diameter of about six to eight inches appears to be a suitable compromise between the conflicting requirements, It would be desirable for the logs to extend the length of the core, but radiation damage may induce a tendency for the logs to bow outward and this would increase the volume fraction of fuel in the core. This increase is controlled and largely eliminated by using logs 24 in. long and stacking these in a vertical position. The ends are mated by means of pins and sockets., Differential shrinkage of the graphite is now accom- modated by a slight rotation about the pins, Fuel salt should permeate the graphite not more than 0.1% by volume. With this penetration, and 10 volume per cent of fuel in the core, about one per cent of the fuel will be in the pores in the graphite. This is probably tolerable, especially if the accessible pores are those near the surface, as seems to be the case. Fuel stagnation in cracks between blocks and in pores in the blocks is closely related to the problem of afterheat, since the fuel so in- _volved is probably not readxly-dralnablg, Means of flushlng the fuel- salt thus retained must be provided'if possible, and if this is not possible, means must be prov1ded of removlng the heat generated in the “core after dralnageo - Temperature rlse and thenmal stress in the reactor vessel must be limited to tolerable. levelso ,leferentlal thermal expansion will leave a ‘gap between the graphite structure and the reactor vessel not less than 1 in. in the radial direction. It will be necessary to provide some flow through this annulus not only to remove the heat generated there, but also to cool the reactor vessel, 1330 ) It will be necessary to orifice the flow channels, or otherwise vary their width systematically in order to distribute fihe flow through the core compatibly with the power density distribution to achieve uni- form temperature rise in gll fuel channels. Moderator In the MSRE (1), the moderator is constructed of square graphite bars measuring 2 inches on the sides and 63 inches long. Channels 0.4 irches deep and 1.2 inches wide are machined into alternate faces of the ‘bars, which are pinned loosely to beams lying across the bottom of the vessel. The channelsroccupy 22.5% of the volume of the matrix. The shape, size, and spacing of the MBRE-moderator elements are not suitable for the MSCR. The pieces need to be larger and the volume of fuel needs to be of the order of 10 percent. The void fraction in a matrix formed by closely packed cylinders of uniform diameter is 0.093. Such a matrix appears to be structurally stable, is easily assembled in a close array, and provides a minimum of contact between individual pieces. This last is important in that the amount of stagnant fuel is probably roughly proportional to the total area of contact. Small varistions in radius, circularity, and straight- ness of the cylinders can be tolerated, and a voidage as low as 104 still be achieved. Higher voidages can be obtained by machining away portions of the surface of the cylinders; in fact, the voidage can be systema- tically varied both radially and axially in this way. The relative positions of the logs are fixed by contact of the unmachined portions at the top and bottom. It should thus be possible to reduce power.peaking in the reactor and, by a combination of orificing and power flattenihg, to obtain a good match between radial distributions of flow and power density and thus achieve approximate equal temperature rise in all‘fuel channels. In the MSCR study, fuel-volume fractions in the range frdm 0.1 to 0.2 were investigated. The optimum fraction appears to be slightly i:} o » 3 331 greater than 0.l. It appears undesirable to design a matrix having a fuel fraction much smaller than this, At low fuel fractions, dimensional tolerances in the machining of the logs and assembling of the matrix in- troduce uncertainties that become an appreciable fraction of the fuel volume, This uncertainty can be reduced somewhat by using logs of large cross section, but there are limits, viz.: (a) a limit on the size of log that can currently or in the foreseeable future be manufactured; (b) a limit imposed'by thermal stress in the log, which becomes excessive with increasing size. Graphite 1ogs measuring 16 inches in diameter can be made now (1962); while these are not of a grade satisfactory for use in the MSCR, it is not a great extrapolation of current technology to postulate the availability of graphite logs of satisfactory grade in sizes up to 8-inches in diameter, and this appears to be as large as thermal strain considerations will allow, I1f the volume fraction of fuel in the core is small, then any slight variation in this volume fraction would have an appreciable effect on the reactivity and might result in power excursions. These variations might arise in any of a number of ways. For instance, radiation damage might result in the accumulation of stresses in the graphite which, upon sudden removal of external restraints (such as by the failure of the hoops, etc.) or by the yielding of the material itself, might result in gross movement of the matrix, and a sudden increase in reactivity, Also, at very low volume fractions, the optimum concentrations of fissile and fertilg,iégtopeé"inthefuel stream becomes high, and this imposes requirements on the désign“of the external system in regard to hold-up, velocitiés;:étc.,'that are_difficult to meet. V,Permeatidn,of Graphite by Salt -_.Graphite,iS*not'impervibus,to salt. The presence of_anfappreciable _fractionibf-the salt'ifi stagnant pockets introduces a nfimber.of problems (suéh'as that associated with the fate of fission products generated in stagnant fuel). The "theoretical" density of graphite is 2.25 grams/cc. MSRE graphite has a bulk density of about 1.83, and thus the pore 332 fraction is about 0.16. The pores accessible to a wetting fluid, such as kerosene or a gas, however, amount to only'0.07% of the volume. Graphite is not wetted by molten flyoride salts, and the penetration is an order less than that of a wetting fluid. Treatment by any one of" several methods (5) reduces the penetration further, either by filling the pores, by closing them, or by making the entrances smaller, Tests in which CGB-X graphite, newly proposed for the MSRE, was exposed to salt at 1300°F and 150 psi for 100 hours resulted in volume fraction permea- tions of 0,001 and 0,0002 (2, p. 93). The occlusions of salt lay mostly at the surface, and were presumably in reasonably good diffusional com- munication with the bulk fluid. The pieces tested were necessarily | small, and the proportion of surface exposed was high. Estimates based on expected frequency of surface pores in larger pieces led to a predic? tion of a penetration not greater than 0.0016 in the MSRE graphite at 65 psi pressure in the salt (2, p. 91). Taking advantage of the fact that the number of surface pores can be reduced by proper orientation of the surface grains and that with larger bars the ratio of surface~to- volume is less, it appears plausible that a salt-accessible pore fraction of 0,001 can be assumed for MSCR graphite. Now, if the graphite occupies 90% of the matrix, salt-accessible pores in the graphite amount to roughly 0.1%, which is only 1% of the total fuel fraction. However, the fuel in these pores may be retained when the reactor is drained, and this may present a serious operational difficulty in regard to cooling the reactor after shutdown, especially if the core is drained shortly after operation at power. Graphite Shrinkage The graphite will, of course, be subjected to radiation damage, mostly from fast neutrons. At the temperatures anticipated in the MSCR, the graphite will shrink. Since that side of a log closest to the center of the core will absorb more radiation than the outer side, the logs will tend to bow outward, and increase the volume fraction of the fuel. These effects will take place slowly, of course, and can easily be e 333 compensated, in respect to criticality, by increasing the concentration of thorium in the fuel or by decreasing that of the uranium, The breeding ratio will necessarily decrease, however, due to shifting of the C/Th/U ratio from the optimum. As mentioned above, the effect is minimized by the use of short logs. Graphite Replacement The precise effect of graphite shrinkage on the performance (as mea- sured by the fuel cycle cost) has not been determined; however, if it proves to be serious, one or more of several countermeasures may be taken, The simplest would be to replace the graphite periodically. The excess sinking fund cost (@ 6.75%) over that corresponding to sinking fund amor- tization over a thirty-year life (1.11%, and which is charged off to cap- ital costs) is listed below in Table 0.1 for several replacements periods. Table 0.1l. Power Cost Increment for Replacement of Graphite® Moderator in MSCR Replacement Period Incremental Power Cost (years) (mills/kwhr) 0.58 | - 0,096 0 0.037 s 0,018 20 R 0,008 #@ $6/1b. - It is seen that, if the replacement occurs no oftener than once in tén years,fthe.incremental cost is tolerable, being less than IQ% of a typical fuel cost of 0.75 milié/kwhr. - If the shrinkage rate is suchfaS'to.require,replacement_more often than this, then part of the fuel volume fraction increase might be avoided by restraining the core and preventing the bowing of the logé. Hoops of 334 molydbenum, which have a coefficiént of thermal expansion very.nearly : equal to that of graphite (3) could be placed around the core (%, p. 32). - Of course, should the hoops fail suddenly, an excursion in the power level might result. On the other hand, if the hoops held, accumulated stresses in the graphite might result in the formation of cracks in which the fuel might stagnate with deleterious effect. | The fuel volume fraction can be made self-preserving in spite of radiation induced shrinkage by use of interlocking blocks of graphite. The moderator elements are cubes having four holes, or sockets, in the four quadrants of thé upper face and four corresponding pegs extending from the lower face. After a laYer of blocks has been laid, the blocks of the next layer are positioned so that the axis of each block lies over the intersectibns’of the fuel chafinei planes bgtween'adjacent blocks.in the lower layer, with its pins fitting into sockets in four.blocks'in that layer. Thus each cube is pinned to four overlapping cubes in the . layer above and four in the layer below. With cubes measuring 8 inches along an edge, and a void fraction of 10%, the thickness of the passage between adjacent cubes is approximately 0.4 inches when the blocks rest directly on the blocks below, and 0.3 inches when uniform clearance is provided on all six sides of the blocks. Now, as the blocks shrink, the fuel volume fraction is invariant, since this is’ determined by the spacing of the pegs and sockets and the dimensions of the blocks. This arrangement, while solving one problem, introduces others, chief of which is a much increased resistance to flow of the fuel through the core, which may increase by a factor of perhaps 20. This would in- crease the pumping cost and the design requirements for pumps, heat exchangers and reactor vessel, _ If difficulty with short-circuiting of the fuel through the annular gap between moderator and reactor vessel is encountered, this could be controlled by eliminating the gaps between the outer blocks of graphite so that the blocks fit tightly one against the other., The annulus thus becomes a channel unconnected to the voids in the moderator, and the flow through it could be orificed at the top where the pile floats up against the upper support grid. b i b1 A SHSREARS1 " L 335 The foregoing discussion of the problem of graphite shrinkage indi- cates some solutions that might be applied if the problem should prove to be serious, but it should not be inferred therefrom that the problem is known to be serious, for in fact, it may not be. There are indica- tions (6) that radiation damage may anneal and saturate at the tempera- ture of operation. In that event, the core would be designed so that, after the steady state is reached, the fuel volume fraction is at the desired value, Differential Expansion The coefficient of thermal expansion of graphite is smaller than that of INOR (compare Tables 3.2 and 3,1)., Thus, in.a 20?ft core with the graphite just filling the vessel at room temperature, there will be a gap about 2/3-in, wide between the graphite and the vessel wall at 1100°F, If further allowance is made for dimensional tolerances between metal and graphite during assembly, the gap cannot be much less than one inch thick at operating temperatures., This gap will contain fuel salt, and this fact must be taken into account in evaluating the performance of the reactor and in the design of the core and the reactor vessel. In a 20 ft cylindrical core the volume of fuel in the annulus amounts to ~ 100 fts, which is an appreciable fraction of the volume of fuel in the matrix (v 600 £t3), This fuel lies in a region of low gegfron population so that it adds little to the re- activity. Also, it is nearly opaque to thermal neutrons which are absorbed,,multipliéd byréta,.and re;emitted as fast'néufrons,'thus:in- - creasing;the leakage. This concentrated source of fast neutrons and concomitant-gammarradiation adjacent to the reactor vessel wall intro- duces desigh problems_° It mayffié néceésafy to pfovide a:thermal shield between the fuel anfiulus and the wall. The most suitable material for this shield is INOR, but this must be cooled. The only available cooiing medium in this Siffiation'is the fuel sait,ran& its use for this purpose further increases the nonactive inventory of valuable materials. 336 It may be possible to use the fuel annulus as a downcomer for fuel coming from the heat exchangers, This may or may not result in some - saving in fuel inventory, depending on the location of the exchangers and their requirements for draining, etc. The annulus is so used in the MSRE (1), References 1. A, L. Boch et al., The Molten—Salf Reactor Experiment, Power Reactor 2, 3. 4, 5. Experiments, Vol. I, pp. 247-292, International Atomic Energy Agency, Vienna, 1962, Oak Ridge National Laboratory, MSRP Semiann. Progr.;Rept. Feb. 28, 1962, USAEC Report ORNL-3282, ' S. E. Beall et al., Molten-Salt Reactor Experiment Preliminary Hazards Report, USAEC Report ORNL CF-61-2-46 (with Addenda 1 and 2), Oak Ridge National Labo.:atory, February 1961. Oak Ridge National Laboratory, MSRP Quarterly Progr. Rept July 31, 1960, USAEC Report ORNL-301u, J. W, H., Simmons, The Effects of Irradiation on the Mechanical Proper- ties of Graphite, Proceedings of the Third Conference on Carbon, Sections E and F, Pergamon Press, 1959. e, v 1. 11. gi) 13. -t 337 BIBLIOGRAPHY L. G. Alexander et al., Nuclear Characteristics of Spherical, Homo- geneous, Two-Region, Molten-Fluoride-Salt Reactors, USAEC Report ORNL-2751, Oak Ridge National Laboratory, September 1959. L. G. Alexander et al., Experimental Molten-Salt-Fueled 30 Mw Power Reactor, USAEC Report ORNL-2795, Oak Ridge National Laboratory, March 1960. L. G. Alexander et al., Thorium Breeder Reactor Evaluation — Part I — Fuel Yield and Fuel Cycle Costs in Five Thermal Breeders, USAEC Report ORNL-CF-61-3-9, Oak Ridge National Laboratory, March 1961, and USAEC Report ORNL-CF-61-3-9 (Appendices), Oak Ridge National Laboratory, March 1961. 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F;'Todd,_Uranlum-235 Fission Product Production, USAEC Report ORNL-2127, Oak Ridge National Laboratory, November 1958. 'C. M Blood, Solubility and Stability of Structural Metal Difluorides “in Molten Fluoride Mixtures, USAEC Report ORNL-CF-61-5-4, Oak Ridge National Laboratory, May 1961. A. L. Boch et al. The MDlten-Salt Reactor Experiment, Pcwer Reactor - Experiments, Vol. I, Internatlonal Atomic Energy Agency, Vienna, 1962, pp. 247-292. | J. 0. Bradfute et al., An Evaluation of Mercury Cooled Breeder Reactors, USAEC Report ATL-A-102, Advanced Technology Laboratory, October 1959. 14. 15. 16. 17. 18. 19. 20. 21, 22. 23. 24. 25. 26. 27. 338 R. C. Briant and A. M. Weinberg, Molten Fluorides as Power Reactor Fuels, Nuc. Sci. Eng., 2:797 (November 1957). J. Bulmer et al., Fused Salt Fast Breeder, USAEC Report ORNL-CF-56- 8-204 (Del.), Oak Ridge National Laboratory, August 1956. | D. 0. Campbell and G. I. Cathers, Processing of Molten Salt Power Reactor Fuels, Ind. Eng. 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MacPherson, and F. Maslan (editors), Fluid Fuel Reactors, Addison-Wesley Publishing Company, Reading, Massachusetts, 1958. H. G. MacPherson, Molten Salts for Civilian Power, USAEC Report ORNL-CF-57-10-41, Oak Ridge National Laboratory, October 1957. H. G. MacPherson et al., A Preliminary Study of Molten Salt Power Reactors, USAEC Report CF-57-4-27 (Rev., Del.), Oak Ridge National Laboratory, April 1957. H. G. MacPherson et al., Molten Salt Reactor Program Status Report, USAEC Report ORNL-2634, Oak Ridge National Laboratory, November 1958. H. G. MacPherson et al., A Preliminary Study of a Graphite-Moderated Molten Salt Power Reactor, USAEC Report CF-59-1-26, Oak Ridge National Laboratory, January 1959. H. G. MacPherson, Optimizing the Molten-Salt Reactor for Minimum Doubling Time, p. 335 of Proceedings of the Conference on the Physics of Breeding, USAEC Report ANL-6122, Argonne National Laboratory, October 1959. H. G. MacPherson, Molten-Salt Breeder Reactors, USAEC Report ORNL-CF- 59-12-64 (Revised), Oak Ridge National Laboratory, December 1959. H. G. MacPherson, Molten-Salt Reactors:’ Report for 1960 Ten-Year Plan Evaluation, USAEC Report ORNL-CF-60-6- 97, Oak Ridge National Labora- tory, June 1960. B. Manowitz, Fuel Reprocessing Costs, Nucleonics, 20(2):60, February 1962 . W. B. McDonald and C. I. McGlothlan, Remote Maintenance of Molten Salt Reactors, USAEC Report ORNL-2981, Osk Ridge National Laboratory (in preparation September 1962). | R. P. Milford et al., Recovering Uranium Submarine Reacfior Tuels, Ind. Eng. Chem., 53:357, May 1961. J. W. Miller, Thorium Resonance Cross-Sections for Thermal Breeder Reactor Study, USAEC Report ORNL -CF-61-1-26, Oak Rldge National Lab- oratory, January 1961. - E. C. Moncrief, Corrosion of the Volatility Pilot Plant INOR-8 Hydro-r - fluorinator and L-Nickel Fluorinator after 21 Nonradioactive Dissolu- ‘tion Funs, USAEC Report (RNL-TM-186, Oak Ridge National Laboratory, ' March 1962 | M. S. Mbore, MIR Nuclear Phys1cs Group, personal communication to C. W. Nestor, Oak Ridge National Laboratcry, March 1960. 70. 71l. T2 73. T . 75, 76. 77. 781 79. 80. 81. 82. 342 J. P. Murray et al., Economics of Unirradiated Processing Phases of Uranium Fuel Cycles, Proceedings of the Second International Con- ference on the Peaceful Uses of Atomic Energy, Geneva, 1958, Vol. 13, - Paper No. Pp439, p. 582, United Nations, New York, 1958. E. A. Ne hewé Thermal -and Resonance Absorption Cross Section of the 233y, 235y, 23%py Fission Products, USAEC Report ORNL-2869, QOak Ridge Natlonal Laboratory, March 1960. C. W. Nestor, Multigroup Neutron Cross Sections, USAEC Report ORNL-CF- 61-6-87 (Rev.), Oak Ridge National Laboratory, June 1961. Oak Ridge National Laboratory, Chemical Technology Division Unit Opera- tions Section Monthly Progress Report, USAEC Report ORNL-TM-34, p. 44 (7.0 Volatility) Oak Ridge National Laboratory, June 1961. Osk Ridge National Leboratory, Reactor Chemistry Division Annual Pro- gress Report for Period Endlng January 21, 1951, USAEC Report ORNL- 3127. Oak Ridge National Labbratory, Molten-Salt Reactor Program Quarterly Progress Report for Period Ending October 31, 1957, USAEC Report ORNL-2431. Oak Ridge National Laboratory, Molten-Salt Reactor Program Quarterly Progress Report for Period Ending June 31, 1958, USAEC Report ORNL- 2551. Ozk Ridge National Laboratory, Molten-Salt Reactor Program Quarterly Progress Report for Period Ending October 31, 1958, USAEC Report ORNL-2626. Ozk Ridge National Laboratory, Molten-Salt Reactor Project Quarterly Progress Report for Period Ending January 31, 1959, USAEC Report ORNL-2684%. Oak Ridge National Laboratory, Molten-Salt Reactor Project Quarterly Progress Report for Period Ending Aprll 30, 1959, USAEC Report ORNL- 2723, Osk Ridge National Laboratory, Molten-Salt Reactor Program Quarterly- Progress Report for Period Ending July 31, 1959, USAEC Report ORNL- 2799. Oak Ridge National Laboratory, Molten-Salt Reactor Program Quarterly Progress Report for Period Endlng October 31, 1959, USAEC Report ORNL-2890. O2k Ridge National Laboratory, Molten-Salt Reactor Program Quarterly Progress Report for Periods Ending January 31 and Aprll 30, 1960, USAEC Report ORNL-2973. o "% 83. 84. 85. 86. 87. 88. 89. 20. 91. 92. 93. 9. : 95: %. 343 QOak Ridge National Laboratory, Molten-Salt Reactor Program Quarterly Progress Report for Period Ending July 31, 1960, USAEC Report ORNL- 3014. Oak Ridge National Laboratory, Molten-Salt Reactor Program Progress Report for Period from August 1, 1960 to February 28, 1961, USAEC Report ORNL-3122. Ozk Ridge National Laboratory, Molten-Salt Reactor Program Progress Report for Period from March 1 to August 31, 1961, USAEC Report ORNL- 3215. Osk Ridge National lLaboratory, Molten-Salt Reactor Program Semiannual Progress Report for Period Ending February 28, 1962, USAEC Report ORNL-3282. Ozk Ridge National Laboratory, Molten-Salt Reactor Program Semiannual Progress Report for Period Ending August 31, 1962, USAEC Report ORNL- 3369. Organization for European Economic Cooperation, Nuclear Graphite, OEEC Dragon Project, Proceedings of the Symposium Held at Bournemouth, G.B. Europ. Nuclear Energy Agency, p. 259 Session I-V, NP-11221, November 1959. R. G Orrison, Union Carbide Nuclear Company (Y-12 Plant), personal communication to L. G. Alexander, Oak Ridge National Laboratory, December 1959. J. J. Pattenden, Fission Product Poisoning Data, USAEC Report ORNL- 2778, Oak Ridge National Laboratory, October 1959. W. D. Powers and G. C. Blalock, Enthalpies and Heat Capacities of Solid and Molten Fluoride Mixtures, USAEC Report ORNL-1956, Oak Ridge - National ILaboratory, January 1956. C. C. Randall et al.,VStudy of:Mercury Binary Cycles for Nuclear Power Plants, Report WCAP-1832, Westinghouse Electric Corp., July 1961. J. E. Ricci, Guide to the Phase Diagrams of the Fluoride Systems, USAEC Report ORNL-2396, Ok Ridge National Laboratory, November 1958. Sargent and ILundy, Enginéers,'Power Cost Normalization Studies, Civilian Power Reactor Program — 1959, USAEC Report SL-1674, January “1960. : Sargent and Lundy,,Engineers, Capital Cost Evaluation 1000 Mwe Molten Salt Converter Reactor Power Plants, USAEC Report SL-1954, June 1962. Sargent and Lundy, Engineers, Capital Investment for 1000 Mwe Molten Salt Converter Reference Design Power Reactor, USAEC Report SL-1994, December 1962. 97. 98. 990 100. 101. 102. 103. 104. 105. 106. 107. 108. 109. 344 J. H. Shaffer, W. R. Grimes, and G. M. Watson, Boron Trifluoride as a Soluble Poison in Molten. Salt Reactor Fuels,: Nuclear Sci. Eng., 12:337, ‘March 1962. J. W. H. Simmons, The Effects of Irradiation on the Mechanical Pro- perties of Graphite, in Proceedings of the Third Conference on Carbon, Sections E and F, Pergamon Press, 1959. : I. Spiewak and L. F. Parsly, Evaluation of External Hold-up of Circu- lating Fuel Thermal Breeders as Related to Cost and Feasibility, USAEC Report ORNL-CF-60-5-93, Oak Ridge National Laboratory, May 1960. P. E. Spivak et al., Measurement of Eta for 233U, 235U, and *3°Pu with Epithermal Neutrons, J. Nuc. Energy, 4:70, January 1957. R. E. Thoma, Crystallization Reactions in the Mixture LiF-BeF,-ThF, (67.5-17.5~15 mole %) BELT-15, USAEC Report ORNL-CF-59-4-49, Osk Ridge National Laborstory, April 1959. R. E. Thoma (Editor) Phase Diagrams of Nuclear Reactor Materials, USAEC Report ORNL-2548, Oak Ridge National Laboratory, November 1959. R. E. Thoma and W. R. Grimes, Phase Equilibrium Diagrams for Fused Salt Systems, USAEC Report ORNL-2295 (Decl.), Oak Ridge National Laboratory, June 1959. USAEC, Summary Report: AEC Reference Fuel Processing Plant, USAEC Report WASH-743, October 1957. ' W. T. Ward et al., Rare Earth and Yttrium Fluorides — Solubility Relations in Various Molten NaF-ZrF, and NaF-ZrF;-UF, Solvents, USAEC Report ORNL-2421, Oak Ridge National Laboratory, January 1958. W. T. Ward et al., Solubility Relations Among Rare-Earth Fluorides in Selected Molten Fluoride Solvents, USAEC Report ORNL-2749, Osk Ridge National Laboratory, October 1959. G. M. Watson and R. B. Evans III, Xenon Diffusion in Graphite: Effects of Xenon Absorption in Molten Salt Reactors Containing Graphite, USAEC Report ORNL-CF-61-2-59, Oak Ridge National Laboratory, February 1961. C. F. Weaver et al., Phase Equilibria in Molten Salt Breeder Reactor Fuels, 1 — The System LiF-BeF,-ThF, , USAEC Report 0RNL-2896, Oak Ridge National Laboratory, December 1960. D. B. Wehmeyer et al., Study of a Fused Salt Breeder Reactor for Power Production, USAEC Report ORNL-CF-53-10-25, Oak Ridge National Labora- tory, September 1953. 5 - ™~ e wi 110. 111. 112. 345 C. H. Wescott, Effective Cross SectianValues‘fqr Well-Moderated Thermal Reactor Spectra, Canadian Report CRRP-787, Chalk River, Ontario AECL-670, August 1, 1958 (Rev. 1958).- N. F. Wikner and S. Jaye, Energy-Dependent and Spectrum-Averaged Thermal Cross Sections for the Heavy Elements and Fission Products for Various Temperatures and C:?3°U Atom Ratios, USAEC Report GA- 2113, General Atomic Division, General Dynamics Corporation, June 1961. H. U. Woelk, Molten Salts in Nuclear Technology, Chemie-Ingenieur- Technik, 32:765, 1960. AEC-TR-4774 by A. R. Saunders and H. H. Stone, Oak Ridge National Laboratory, August 1961. > R & O ot 1. 2-20. 21. 22. 23. 24, 25. 26. 27. 28. 29. 30. 31. 32. 33. 34. 35. 36. 37. 38. 39. 40. 4. 42. 43. by o 45. 46. 47, 48. 49. 50. 51. 52. 53. 54.. 55. 56. 57. 58. 59. 60. 61. 62. 63. 64 65. 66. 67. PRGN EOEOR QRN EEOE P B RN O E . * copREmEoEEIsorEmE=EGE ?:?i?>g:;:g>g*z:gzgzgig:5+t*c>c>niujc4c:t+w:¢1a:t*m:q1m,U Anderson Alexander Bauman Beall Bennett Bettis Blankenship Blizard Boch Bohlmann Borkowski Briggs . Bruce Campbell antor Carter Cathers Chapman Carlsmith Collins Cock Craven Culler Delene DeVan Donnelly Douglas Engel Ergen Fraas Frye, Jr. Gall Gift Goeller Grimes Grindell Guthrie Hammond = Hsubenreich . Hise " Hoffman Janney (K-25) Jordan Kasten Kedl Kelley Kerlin Lane Larson - - HPEHCXEE RO TEAEEERED - 3477 Internal.DiStribution 68. C. G. Lawson 69. R. B. Lindauer 70. J. L. Lucius - 71. M, I. Lundin 72. R. N. Lyon 73. H. G. MacPherson 74. H. C. McCurdy 75. W. B. McDeonald 76. H. F. McDuffie 77. R. P. Milford 78. A. J. Miller 79. R. L. Moore 80. J. C. Moyers 8l1. R. W. Qlson 82. H. R. Payne 83. A. M. Perry 84. B. E. Prince 85. R. C. Robertson 86. M. W. Rosenthal 87. H. W. Savage 88. A. W. Savolainen 89. D. Scott 90. J. H. Schaffer 91. M. J. Skinner 92. 1I. Spiewak -93. ‘W. G. Stockdale 94. A. Taboada 95, J. R. Tallackson 96. R. E. Thoma 97. D. B. Trauger 98. J. W. Ullman 99. D. R. Vondy 100. R. Van Winkle 101. G. M. Watson 102. A. M. Weinberg 103. J. H. Westsik 104. G. D. Whitman - 105. K. J. Yost 106. Gale Young ' 107. Biology Library 108-109. Reactor Division Library 110-111. MSRP Director's Office, o Room 219, Building 9204-1 - 112-113. ORNL Y-12 Technical Library, Document Reference Section 114-115. Central Research Library 116~145. Laboratory Records Department 146. Laboratory Records, ORNL, R.C. 147. 148. 149-150. 151. 152. 153. 154. 155. 156. 157-158. 159. 160. 161. 162. 163. 164 . 165. 166. 167. 168. 169. 170. 171-185. 348 External Distribution G. 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