ORNL-TM-1060.txt

maflan of a prehmmary
‘yse at the Ok Ridge Nattonu{

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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

 

 

 

 

 

 

 

 
 

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: - 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

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SEPTEMBER 1965

OAK RIDGE NATIONAL ILABORATORY
Oak Ridge, Tennessee

- Operated by
UNION CARBIDE CORPORATION
for the
U. S. ATOMIC ENERGY COMMISSION

 

 
 

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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.

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- 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
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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

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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

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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 <sveseesssansnon
4.8.13 Coolant Salt Sampling «.esvervsseeveeen. Ceeerueenean
4.8.14 Freeze Valves ...... ceerencaneas Cecoarenne ceseens
4.9 Auxiliary Services and Equipment ............ . oo
4.9.1 Introduction «ceceiesrerestectnariassarasracsssavsens
4.9.2 Helium Cover Gas Supply and Distribution System ....
4.9.3 Reagent Gas Supply and Disposal System ...cocevevnes
4.9.4 Waste Gas System ...cveeeeennnn ceereceecnns ceeesaans
4.9.5 Liquid Waste Disposal System svcecoveecen. ceeseseans
4.9.6 Coolant Pump Lubricating Oil SyStems ««eeeeeeeeeeoos
4.9.7 Preheating System cecceveiieetiaeiritacneresscrianaas
4.9.8 Auxiliary POWEY «oveveccescassnssassnsansanas cereene
4.9.9 Service Water System «vveveeseeerinscessncesanns cees
4.9.10 Control and Station Air Systems ..........0... cos
4.9.11 Cranes and Hoists «....... teseerieneanns . caeeeean
4.9.12 Instrumentation and Control «..ceveeevncrncnes cesens
4.9.13 Plant Utilities cceveevecnnns ceetertecessaanseneanas
4.10 MSCR Design Specificaiions ceceacacsanaas tresereannne
"FUEL PROCESSING «vecocecocvvcoscsoanconnsas sacnas cesesans sesacse
5.1 Reprocessing System .c..ccv0ev... cesesecasssnnasaannnse ces
5.2 Fluoride Volatility Central Plant ....cceveievnessn ceesaa .
5.2.1 Process Design ceeevecencssssatasccsssosas cesans e
5.2.2 Shipping ceceeeccnass ctetsecsesectcncrasasaveans sees
5.2.3 Prefluorination Storage Tanks ..occevvuieniiinnnns
5.2.4 Fluorifia;tor .-I.........‘.............. . sessen oo
5.2.5 CRP Trap and NaF Absorbers ...... cecsesessaraasaas .

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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

 

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CONTENTS (continued)

Captures «cceeee.. cheseanes ceedtesnsiersataaseaeenanes 158

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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

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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

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xiii

CONTENTS (continued)

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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

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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.

 
 

 

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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.

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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

 

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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

 

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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

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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

 
 

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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

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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

 
 

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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

®

 
 

 

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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

 
 

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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.

 
 

 

 

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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

 
 

 

 

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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

 
 

 

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7)

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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)
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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 : |
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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 '/
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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

       
  

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COOLANT

    
  
 

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AN,
TRANDE

L.

 

 

 

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A B

 

 

. e TURBING
TRANMSY.

 

 

:EJ

Fig. 4.5. MSCR Plant General Arrangement. (Sargent and Lundy Drawing)

 

UNAT

TRANLF,

 

 

 

 

 

 

 

 

 
 

 

 

 

  
  
 
   

   

IECULATING WATER
DISCHARGE TUNNEL

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H.P TURBINE
TRANSFORMER

 

 

 

" c-

Fig. 1.6, MSCR Plant — Plan. (Sargent and Lundy Drawing)

   
43

    
    

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.. 908ie’ - $88L0"

TaNES Leal. 984"

SECTION AA CTION "B-&’'

tankAToRd

 

 

S Nes e
50 TOM CRANR L.

L A G

  

ContRo. Roam
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CamLE Room {
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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

 

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? .
r L

 

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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. <Circulation of salt through this'bearing is accomplished
by allowing salt to leak upwards around the shaft and out a small vent
line. A cylindrical casing above the impeller provides expansion volume »
and a volatile fission product purging surface. The surface of salt in
this casing is maintained at the same level as that in the expansion dome | :
under a helium cover gas at 22 psi. This gas overpressure also helps |
prevent impeller cavitation.
Detailed specifications of the MSCR pumps are given in Section 4.10.

4.4.4  Primary Heat Exchanger

The design of the heat exchanger for transferring heat from the fuel
salt to the intermediate coolant salt is shown in Fig. 4.10. This shell
and tube heat exchanger is of a U tube configuration designed particularly
for accessibility to the tube sheet from above without disturbing con-
necting piping. Removal of the tube bundle through the top head is also
possible with the given design. With this configuration, however, the | o
more valuable fuel salt, which normally would be circulated through the
tubes, is put on the shell side to permit ready drainage. . The tube bundle | :
and associated baffles, which can slide into the shell from above, hang
from the tube-sheet which rests on a shelf machined into the shell. A
circumferential seal weld joins the tube sheet to the shell and separates.
the coolant from the fuel salt. A hbld-down ring keeps the tube sheet in
Place in case pressure on the shell side should exceed pressure on the
tube side. Normal design conditions regquire that the coolant salt (tube
side) be kept at a pressure higher than that of the fuel salt.

 
 

49

¢

ORNL-LR-PWG 63-2802

  

Coolant Salt

Wik §_ <

Bl _
mwv_mmmw§§§mk

 

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Shear Tubes
Inor Shell

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Fuel Salt

 

7
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N7,

Drain

-

MSCR Primary Heat Exchanger.

Fig. 4.10.

 

 
 

 

50

The end-closure may be sealed against leakage of coolant salt to the
outside in several ways. A bearing and sealing suiface (much like a
valve seat) is provided at point "A" shown on Fig. 4.10. A frozen salt
seal is maintained in the annulus above "A". A seal weld could be made
at the top between the flange on the shell side and the inverted head.

Or lastly, double-gaskets on circles inside and outside of the seal weld
1lip could confine leakage if the seal weld, or other seals ahead of the
seal weld, should fail. A massive hold-down flange bolted to the top
flange provides the strength necessary to hold the inverted head against
the several hundred pound pressure of the coolant salt.

A partition attached to the inverted heat fits closely into a dia-
metrical slot in the tube sheet to separate the inlet coolant salt plenum
from the outlet plenum. A small amount of by-pass flow may occur thraugh .
the very small gap between the slot and the partition. The ifl#erted head
is shaped so as to minimize the volume of coolent salt in the tube-sheet
distribution plenums. The inlet and outlet coolant nozzles are noncircu-
lar in cross section (flattened and broadened so as to minimize head-space
without sacrificing flow area).

Not shown are cooling and heating provisions for the top head, heat-
ing provisions for the shell or the insulation which must be provided.

In order to avoid excsssive thermal stresses in the tube sheet it may
be necessary to reduce the thermal gradient acroés the tube sheet by pro-
viding insulation between it and the fuel salt. |

An appreciable heel of coolant salt will remain in the tubes after
draining. In order that the salt-containing internals of the heat ex-
changer be exposed only to inert atmospheres when the top closure is re-
moved, the upper portion of the heat exchanger (above the coolant nozzle)
is provided with a gas-tight caisson that extends from the heat exchanger
to the locality of the maintenance equipment.

The specifications given in Section 4.10 are based on eight heat ex-
changers per reactor, each with a heat load of 1.006 x 10° Btu/hr per unit.

 
 

 

 

©

heater.

51

4.5 Intermediate Cooling System

4.5.1 Introduction

Leakage of steam into the fuel salt would result in precipitation of
uranium; therefore, a materisl which has good heat transfer properties and
is compatible with both fuel salt and steam must be interposed between
these two mfiterials. of fihe various possible choices, a barren salt of
LiF-BeF,; appears to be the most promising and is used as a basis for the
present MSCR design invelving a Loeffler steam generation system. In this
design the intermediate cooling system consists of the heat exchanger al-
ready described, coolant salt pumps, steam superheaters and steam reheat-

€rs.

4.5.2 Coolant Salt Pumps

The coolant salt pumps are similar in design to the fuel salt pumps
but have a higher capacity and head.  Eight pumps serve the sixteen super-
heaters and four reheaters and each pump circulates 13,900 gpm of coolant
salt at a temperature of 950°F and the primafy heat exchanger inlet pres-
sure of 300 psia. The developed head of the pump is approximately 270 ft.

Additional specifications for the coolant pumps are given in Section 4.10.

4.5.3 Steam Superheaters

 

Steam from the superheaters is divided among several streams. About
20% is useéd to drive the'tgrbogefierators for the production of electricity;
a much smaller fraction is used to driVe other turbines which power feed-

water pumps and the steam circulators. The balance is mixed with feedwater

in the Loeffler'boiler.td'prOduce'saturated steam required in thérsuper-

A U-tube-in‘U-shéll arrengement with fixed tube sheets and counter-
current flow of coolent salt-and steam is shown in Fig. 4.11l. Coolant
salt is on the shell-side, and steam flows through the tubes. The unit is

mounted on_its side to minimize floor space. The shell-side is unbaffled

" to minimize pressfire drop.' The heat transfer coefficient with near-laminar

flow of coolant salt is about 15% less than for baffled flow, and requires

 
 

SUPERHEATED

STEAM OUT

SATURATED

STEAM IN

 

COOLANT SALT IN

52

 

 

 

COOLANT SALT QUT

Fig. 4.11.

 

 

 

 

 

 

TUBE SHEET

 

 

 

 

 

MSCR Steam Superheater.

 

ORNL-LR—DWG 39169A

 
 

 

 

 

"

»

53

more tubing. To prevent fieézing of the salt, the superheate: is brought
up to operating temperature by steam generated in an auxiliary, oil-fired
boiler. _ |

Coolant salt operating between 950 and 1100°F circulates through the
shell-side. Steam enters at 670°F and 2490 psi, and leaves at 1000°F and
2465 psi.

4.5.4 Steam Reheaters

In order to provide greater energy availability for the turbine work
cycle and to provide lower pressure steam to drive auxiliary turbines, the
high pressure turbine exhaust steam is directed to the reheaters, the
auxiliary turbines, and the second feedwater heaters. The same general
design criteria apply to the reheaters as to the superheaters, although
the steam is at a lower pressure.

As with the superheaters, the reheaters are constructed of INOCR-8
and alloy steel. Eight units provide the desiréd heat transfer surface.
Again, the coolant salt flows in the shell countercurrently to the steam
passing through the tubes.  Comparable pressure drop and heat transfer
conditions exist ih the reheater as in the superheaters, ajthough baffling
(to improve heat transfer) on the shell introduces a pressure drop of

about 40 psi.
4.6 Power Generation System

46,1 Introducfiion”-.

- Before selecting the power generation system for the MSCR a number of

‘alternative gpproaches were considered in some detail. These included the

‘use of (a) a mercury'boiler,'(b)_a "Kinyon" boilef, (c) a "Bettis" boiler,

and (d) a flLoeffler"-boiler, The characteristics of these alternative

~ steam generation systems are summarized as follows.

The use of binary mercury-stéam power cycles for power generation

from solid fuel reactors has been ‘studied by Bradfute etial, (13) and

‘Randall et al. (92). Kinyon and Romie (55) considered application of the

binary cycle to molten salt reactors. The binary system, in addition to

 
 

54

mitigating the thermal stress problem, has the fur£her.adv§n£age'df.a"_
higher thermal efficiency (95). On,ihe other hand, bimetallic.heat ex-
changers are required at thé high temperafiure end, and'it.iéant'dbvious 7
that the fuel salt volume will be small compared to that involved in the
Loeffler system.

The "Kinyon" boiler (53) employs bayonet elements composed of three
concentric tubes. Water is introduced in the annular passage surrounding
the innermost tube and is boiled by heat transferred from superheated
 steam flowing downward in the outer annulus. Vapor is separated from the
liguid at the top of the inner annulus; the liquid is routed down the
central tube to a plenum from which it is recycled to the boiling annulus.

Vapor passes down the outer annulus; it is superheated by salt surrounding

the bayonet element and in turn boils the water rising in the inner an-
nulus. This arrangement results in tolerable thermal stresses in'the_
tubing. . The system avoids the use of steam recirculators.  Furthermore,
the bayonet tube boiler may be more compact and less expensive than the
equivalent superheaters and boilers in the Loeffler system. It is not
clear, however, that the required rate of vapor-ligquid separation can be
achieved with present technology.

The "Bettis" boiler is a modification of the "Lewis" boiler (59,55)
wherein a bayonet tube containing water and steam is inserted into a
thimble extending down into the fuel salt. An inert buffer salt occupies
the annular space between the pressure tube and the thimble, providing
thermal contact along with physical isolation. This system avoids use of
steam circulators and also eliminates the necessity for pumping the in-
termediate salt.

4.6.2 loeffler Boiler System

In the system selected for the reference design, four boilefs gen-~.
erate steam by direct cbnfiact between the steam from the.superheaters and
the feed water from the turbine cycle. Four steam compressors circulate
the saturated steam from the boilers through the superheaters of the inert
salt system, providing steam at 2400 psia, 1000°F at the turbine throttle
and at the superheated steam inlet of the Loefflér boilers. |

C

e

 
 

 

)

55

These boilers, shown in Fig. 4.12, consist of cylindrical drums with
an inside diameter of about 72 in., with distribution pipes for super-
heated steam and feed water located on the vertical centerline and below
the horizontal centerline of each drum. Conventional cyclone steam sepa-
rators, located above the horizontal centerline and arranged in four paral-
lel rows along the inside of the drum, separate the steam from the boiling.
water. A system of scrubber-type steam driers occupies the upper part of
the drum. Tn operation, euperheated steam and feed water are mixed as
they leave the two distribution pipes and boiling occurs. The boiling
water is guided by internal baffles through the cyelone seperators where
the water is removed by centrifugal force. Steam and a small amount of
moisture flow through the scrubbers, where the steam is dried to approxi-
mately 99.7 quality. The drums are fabricated of carbon steel and are
designed for a pressure of 2625 psia. A wall thickness of 7 in. is re-
quired for an inside diameter of 72 in. and a temperature of 650°F. Each
drum is about 100 ft in length and has hemiellipsoidal heads.

4.6.3 Steam Circulators .

- The steam circulators are single;stage'axial compressors suitable
for a steam flow of approximately five million lb/hr'at a discharge pres-
sure of 2480'psia. The circfilators_are driven by steam turbines, using
steam from the cold reheat lines of the main turbine-generator. At design
conditions, each circulator requires a power input of 4750 BHP. Steam
from the»turbines is returned”to_the«eycle through'the feed-water heaters.
One circulator has aLmbtor-dr{Ve;suitable for full flow at design pressure
to be ueed during etart-up’ehutdbwn:' An oil- fired'package boiler, de=~
81gned for a saturated steam flow of 50, 000 lb/hr at 300 psia, is provided
for startup purposes. ' , :

The system is des1gned to operate as four 1/4 capa01ty units; each

unit consists of one Loeffler b01ler, one steam c1rculator and four super-

heaters. Valves are located accordlng to this phllosophy. Interconnec-

tions between varlous unlts are not prov1ded except at the inlet to the

 boilers and the dlscharge from the superheaters.

 
 

14 in. Sch. 160 ORNL DWG. 65-7908
Steam Header -
Scrubbers

     

Cyclone Separators

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  
  

T

0‘_ Superheated Steam Header

I-f - $ 100 ft | 7‘—|

Operating Conditions (one unit): ) _ \
Inlet Water 2 X 10° 1lb/hr @ 545°F, 2500 psi
Inlet Steam 3.1 x 10% 1b/hr @ 1000°F, 2450 psi
Outlet Steam 5.1 X 10% 1b/hr, sat. @ 2400 psi

 

     
   

Water
Distribution
Header

 

 

Superheated Steam Headers

Fig. 4.12. MSCR .Loeffler Boiler.

 

 

 

9%
 

 

 

 

o)

57

4.6.4 Turbogenerator

The turbogenerator"is a standard CC6F-RH unit Operating at 2400 psia
1000°F and producing a gross electrical outputéof 1083 Mw. As shown in
Fig. 4.3 the units consist of three turbines: two double-flow high pres-
sure units and one douole-flow intermediate pressure unit on one shaft
with a 667 Mw net output plus three double-flow low pressure units on
another shaft with a 416 Mw net output.

The plant auxiliaries have a maximum coincidental loading of 45,000
kws thls power results in a net station output of 1038 Mwe to the station
main power transformers.

Other characteristics of the power generation system are given in

Fig. 4.3 and summarized in Sectlon 4.10.
4.7 Reactor Control System

4.7.1 VIntroduction

The. spec1f1catlons for the MSCR control system depend on the unique
propertles of its fluid-fuel. The main mode of shim control is by adding
UF, to increase reactivity and by replacement of fuel salt with a salt
containing no fissile 1sotopes to decrease reactivity. This control is
supplemented by Operatlonal control with BF3 gas, which is sufficiently
soluble that adequate concentrations may be obtalned in the salt by regu-
lating the partlal pressure of BF3 over the salt surface in the dome.

Fuel is added to compensate burnup and flSSlon product accumulatlon.

VGreater react1v1ty control is’ requlred durlng startup and shutdown The

rstartup procedure 1nvolves a gradual increase in power, durlng whlch the

rising xenon p01son1ng (controlled by the flux level) is compensated by
the addition of new fuel. After shutdown, the reactor is sub-crltlcal
for forty hours whlle @35Xe grows 1n Ir longer shutdown is requlred

then fifty cub1c feet of fuel salt must be removed and replaced by non-

B fissile salt

Emergency shutdown, such as that requlred in a loss-of flow a001dent

is provided by injecting BF3; gas through tubes opening into the core just

~above the bottom support grid. The BF; will displace fuel, will rapidly

 
 

58

dissolve in the salt, and will diffuse into the graphite. All of these
processes will result in a rapid and large decrease in reactivity.

)

4.7.2 Shim Control

- For shim control by fuel addition or removal,.two reactivity-control
drain tanks of 50 cubic ft capacity each and a shim-control salt-addition
tank of 50 cubic ft capacity afé located in the reactor éontainment cell.
All tanks are provided with vents to the reactor dome, gas connections
to the helium pressurizing sjstem,,and Several 100 kw electric heaters
for initial véssel héating and maintaining stored salt tempefature-' The
addition tank is essentially in parallel with the 2 cubic £t salt-addition
metering teank. The drain tank is connected to the reactor_vessei upper
plenum and is installed with respect'to the reactor dome 1iQuid level so

that overfilling will not be possible. Freeze valves are used to isolate

the tanks from the reactor. The reactivity control drain tanks must have,

in addition to heating coils, decay heat removal facilities similar to the
fuel salt drain tanks. They have an independent steam condenser and heat
removal system which will be 1/20 the size of the fuel drain tenk heat

removal system.

4.7.3 Emergency Control

The BF; addition system comprises pressure cylinders manifolded to a
heater. BFs3 flow control and quick-acting injection valves are incorpo=-.-
rated in the supply line connected to the bottom plenum of the reactor
vessel. The flow control facility is operated from the. shim control in-

strumentation and the quick-acting injection system is operated from the

"scram" circuits.

The BF3 gas is stored in six 330 cubic ft, 2000 psi cylindérs located
outside and adjacent to the east wall of the reactor building. The sPeci-
fications for the reactivity-control drain tanks and the reactivity-control

salt-addition tank are givenzin'Section 4.10.

o

 
-

59

4.8 Salt Handling Systems

 

4.8.1 Introduction

Facilities for handling both fresh and spent fuel salt and inter-

mediate coolant salt constitute the heart of the MSCR complex. The func-

tions which must be performed include melting, purifying, charging, re-

moval, storage, and sampling of the salt. These are accomplished by means

of the following systems and facilities shown in Fig. 4.13.

o o

om0 o0

1
.

k.

Fuel salt preparation

Coolant_salt preparation

Reactor salt purification

Coolant salt purification

Reactor salt charging system

UF,; addition facility

Fuel salt drain and storage system
Coolant salt drain system

Spent fuel withdrawal system

High level radiocactive sampling

Coolant salt sampling

4.8.2 Fuel Salt Preparation

 

Fuel salt is received in solid form and must be liquified prior to

processing. Barren salf, having the same composition as the intermediate

coolant salt, i.e., 66% LiF and 34% BeFp, with a liquidus temperature of

851°F is used for preoperational testing and system flushing before and

after maintenance activity. A solid nmixture having this composition is

charged initially. Sufficient LiF and ThF, are added to form composition
68% LiF, 23% BeF, and 9% ThF,. All mixtures are processed through the

same equipment. After routine operation is established, flush salt is

the base material for the lithium/thorium addition process. In this man-

ner, flush salt contaminated by fission products may be consumed and

special treatment for removal of fission products will be unnecessary.

A 12-ft-high, 150 cubic ft tank enclosed by an 8-ft-high, 150 kw

furnace receives solid salt for melting. The furnace, located in the

 

 

 

 

 
 

60

 

 

 

 

. ORNL-DWG. 65-7909
LiF_BeF; LLF Th Fadioaclve Salt Fortable
Sampler C}:’/;:t/;’
Dry Chemucal - ASIEET Pry Chemical
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rom Flask rfi?.,ff r Tan# s ||Teatmet  2-7504(7
Reagent Cas 5-35f(% Decay S 150K °"§‘ ¥
Supply Séorage Tanks Con fo o
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8¢ - 50F¢ ’ Fuel "Salt of‘mgcnt Gag Supply
Ovamn Tanks
tobinun
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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.

 

 
 

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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 Neptuniu޴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 <t NN NNOYWO-0ONETNNEFOMNODOOO0OO0O0O AT HA
TN OU-00O AN NN OO WOOWOOAANRTNANANOOO A A A~~~
NANANNNNOO OO M
n A A A A A A~
2 | __
O | A NN OO OHNMNTVNOROAO N M~ N O D™
& A A A AA A A ANNNNNN AN
OO MNO N M~ .
ANANNANAMNONM

 

 

 
 

 

190

Table A.12 (continued)

 

tr

'R

Edq,

Group

 

2.991 x 10

1.316 x 1071

28
29
30

- 2769

2.991 x 10

3.376

737.0

764.0

1.438 x 10°1

2.99]1 x 10
- 2.991 x 10

1.630 x 107

1.885 x 10™1

31
32

'7.204
8374
18.943

- 2.044 x 1071

2.991 x 10

7.486 x 10

2.25 x 1072
2.25 x 1071

33

1.32 x 10°

34

 

Table A.13.

Uranium-233

 

I>

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; <t
ijel o 1 E

 

By P
N, _ = 13513

13E T T . =
i] A (T, + T)
[1 — e ijt e E ]

l-—c¢€

(9)

The rate of decay of precursors in the external loops is

| ( = ich)( -—7\13TE)
v ags _ Pig¥sesPE) e " J|t—e

= - — (10)
E'E T ,
| E B - [1 _ e—')\ij:(Tc +TE)]

 

 

rklaTE

'The total rate of precursor decay (at equlllbrlum) is 5 j 5 fd¢f Thus

| the fractlon of the delayed neutrons whlch appear in external 100ps is

-TEf TV |1 < ® | o
5 N e , . (11)
i3 J :E'J [1 ~ e—?\ij.(Tc + TE)]

‘“fi fcvc-)\lJTE

' N.. TV '.'.f ,:  . - L
S ' (l - e AlJ c)(l -2 Kij E)

 
 

228

 

TE
=5 Eq. (11) becomes
¢ o .
=L T =N 4T
PijE (l -e c) (l —e E)

—7\13 ( T, + Tp )]

. | - (12)
Aich [l — e

 

Bi;

For using these results in an equilibrium reactor code such as ERC-51

the term.vjzfj¢ pay be replaced by a neutron production rate given by

NCV,
Jd Jd J

where

Cg = reaction rate coefficient for fissions in material j.

Using this in Eq. (12) gives the following result for the number of neutrons

lost in the external loops per neutron produced:

 

N

nyefv; ) H -

T T [1 . e—')\ij(Tc + TE)]
ij e

losses = (13)

The necessary constants for ?3%Th, 233y, 235y, 238y, and 22%Pu are:?

 

 

 

 

 

 

Group ')\i(sec-l) 232qn 233y 2353 238U ' 239py

1 0.0128 0.00085 0.00020 0.0003 0.00015 0.0001

2 0.0315 0.0035 0.00075 0.0018 0.0017 -0.0006

3 0.125 0.0045 0.00105 0.0022 0.0028 0.00045
4 1 0.325 0.0120 0.00075 0.0023 0.0071  0.00085
5 1.55 0.0045 0.00025 0.0007 0.0042 0.0003

6 4.5 0.0009 0.0002 0.0015

0.02625 0.0030 0.0075 0.0023

0.01745

 
 

 

lo

2.

229

References

L. G. Alexander, ERC-5 Program for Computing the Equilibrium States
of Two-Region, Thorium Breeder Reactors, ORNL-CF-60-10-87 (October 20,

1960).

A. M. Weinberg and E. P. Wigner, The Physical Theory of Neutron Chain
Reactors, p. 136, The University of Chicago Press, Chicago, 1958.

 
 

 

230
Appendix G

TREATMENT OF XENON ABSORPTION IN GRAPHITE

L. G. Alexander
Introduction

Some 135Xe is formed directly during fission; hdwever, the major
part (>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+<H,0 for:which a reasonably accurate

price can be computed. However, ai uncertainty exists in the charge for
converting the hydroxide to an anhydrous fluoride predfict since this
operation has not been attempted except in small-scale batehee; this™
charge was estimated. A schedule of basic charges was recommended.(3,4)

for computing the price of lithium fluoride in 99,995 at, % 7Li:

th

 
 

 

 

 

 

 

4}

.

271

L10H0H20 =

LiOH-HQO ~ LiF conversion =
Cascade feed cost () =
Base cost =

Plus 15% AEC overhead (u4)

The values have not been released by the Atomic Energy Commission as
an official price for high purity 7Li; they are confidential information
for internal use only. |

The official price was $120/kg of Li metal as the fluoride in
November 1959 (5) for a grade containing 99.99% of "Li. This price was

adopted for material containing 99.995% Li produced in large quantities.

Recommended Values for Molten Salt Fuel

Based on the values given in Tables J.l, J.3, J.4, and J.5, the fol-
lowing prices are recommended for use in molten salt reactor fuel cycle
calculations. These prices were chosen more or less arbitrarily from the
accumulated data since there was no reason to have more credence in one

value than another.

.~ 8/1b of $/kg of metal

 

Element .  fluoride compound  (as fluoride)
Thorium .50 | 19,00
Beryllium 7,00 80,42
Lithium 170 120,00

Zirgonium' . 4,00 . 16,13

 

&fi; ¥This information is classified.

 
 

 

1.

2,

Je

4,

5,

272
References

R. G, Orrison, Thorium Metal Processes, Y-1312 (December 18, 1959),

Personal communications with E. E, Keller, AEC-ORO, Neal Dow, UCC-
Y-12 Plant, and Murray Hanig, UCC~ORGDP.

Letter from R, G, Jordan (Y-12) to C. A. Keller (AEC-OR0O), No.
Y-A0-2314, January 10, 1962.

Personal communication with E. L. Keller, AEC-ORO (January 18, 1962).

Nucleonies, 17(11): 31 (November 1959).

 
 

 

"

w)

@)

"

‘ifi

273

Appendix K

MSCR POWER LIMITATION RESULTING FROM
MODERATOR THERMAL STRESS

R. H. Chapman

Summarx

In order to achieve the maximum perfdrmance from the MSCR, it is
necessary to operate at a power level corresponding to some limiting con-
dition. The limiting condition may be arbitrarily imposed upon the system
or it may be an inherent feature in the design. In certain conceivable
situations thermal stress in the graphite moderator may limit the power
output of a single reactor core. It is therefore of interest to estimate
the maximum power output as a function of core diameter for the case where
thermal stress is the limiting condition.

It is assumed for fihe purpose of this memorandum that the reactor core
is a cylinder of L/D = 1.0. Honeycomb shaped fuel channels are formed by
the proper spacing of hexagonal graphite moderator prisms. For the purpose
of estimating the thermal stress, it is also assumed that the hexagonal
prisms can be approximated by a right circular cylinder of a diameter equal
to the distance across the flats of the hex. It is noted from the geometry
of the hexagonal unit cell that the fuel volume fraction, flv, the fuel
channel thlckness, t o? ‘and the distance across flats, dc, are interrelated.
Flgure K.1 shows the relatlonshlp for the region of interest. The size
of the unit cell is obtained by adding the channel thickness to the dis-
tance across flats of the mdderator prisms. |

Assuming unlform heat generation, the maximum thermal stress (at the

_surface) is glvenl for a solid cylinder as

o - /7// 2
10E g T,

-¢t= e —— (1)
21— 4k
where
o, = tangential thermal stress, psi,

 
 

 

274

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\ \\ \ \ /l\um Cell
FIMRRRINNY AN
AU Sedx
AN NN orephte """
TR
AU
RANMANRONANNNN
RO RN
N\ N AN
N OO AN
S NMANANR OO
NN \\\\‘\i\\\\\\'-sfi
SRR NN \\\\\\\\ o B
6 \\ '\‘\‘\\\‘\\\Wg
N N R
4 \\‘ \\ \\\\\225
2\ \\ -\\ 04 o
1 —— o

 

 

Fy= FUEL VOLUME FRACTION

Fig. K.1. Hexagonal Unit Cell Relationship.

O

 
 

[}

w)

&

»

275

 

a = coefficient of thermal expansion, E%%L?f ,
E = modulus of elastiéity, psi,
v = Poisson's ratio,
q’/’/ = volumetric heat source, i ’
hr ft°
r, = radius of cylinder, ft,
k = thermalvconductivity, E;E%%"?fi .

It is assumed that the uniform volumetric heat generation term can be re-

lated to the core heat production rate by

£ . q [P
g/’ = L& _%gfi c, (2)
VM P
where
fip = fraction of total heat which is produced in moderator,
QC = core heat production rate, Mwt,
Vy = volume moderator in core, ft°,
_ . B ¢  BTU
Cl = conversion factor__ 3.413 x 10 T =Y ?
max . | | .
—— = ratio of peak-to-average core power density.
P - o _ _

The core moderator volume is given for L/D, = 1.0 as

o T D2 |
Vy = (1=t 4" . | (3)

 

Substituting Eq. (2)_and'(35'into (1) and rearranging, the following ex-

pression is obtained:

 
 

 

276

 

Q 2k 1 — vV P D 3 |
C ¢ i
= — oy | — | — (4)
1 —-fv Cl oE - Bmax fp-ro

Assuming values for the various properties of graphipe, i.e., Oy s O E, v,
and k, and representative values for the fraction of total heat generated
in the graphite and for the peak-to-average power density ratio, a family
of curves of heat production versus diameter may be plotted wherein T, is
the parameter. 7

Choosing an allowable stress value of graphite is a.gomewhat arbitrary
bperation. Experience has shown in the production of AGOT graphite for
~ the EGCR (16 in. X 16 in. X 20 ft long columns) that the mechanical prop-
erties vary widely between different blocks of the material, vary across
the cross-section of the large blocks, and depend on the orientation rela-
tive to the direction of extrusion.? Fracture strain is probably a better
criteria for failure than stress, since it has been shown to be fairly
constant at about 0.1 to 0.2%. Tests have also shown that strength is not
temperature dependent at least up to 1100°F. Since the ratio of stress to
modulus of elasticity appears in Eq. (4), the value of the fracture strain
can be substituted and thus side-step the issue of fraction stress. A
value of 0.1% is assumed as a failure criterion.

For Poisson's ratio a value of 0.4 is recommended, and for the thermal
coefficient of expansion, a value of 2.7 X 10%6 in./in. °F is used.’ A
value of 15 BTU/hr-ft'°F is used for the thermal conductivity at 1200°F.%
A value of 2.5 is assumed for the ratio of peak-to-average power density
to account for the maximum thermsal stress condition. This is about the
value computed for the MSRE.®> 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),
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‘March 1962.

J. W. H. Simmons, The Effects of Irradiation on the Mechanical Pro-
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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,
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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
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G. M. Watson and R. B. Evans III, Xenon Diffusion in Graphite:

Effects of Xenon Absorption in Molten Salt Reactors Containing
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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
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5

 
 

 

 

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™~

e

wi

110.

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112.

345

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Thermal Reactor Spectra, Canadian Report CRRP-787, Chalk River,
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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.

 
 

 

 

 

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Blizard
Boch
Bohlmann
Borkowski
Briggs

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Campbell
antor
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Chapman
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Collins
Cock
Craven
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Ergen
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Frye, Jr.
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Goeller
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Grindell
Guthrie
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Jordan
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- 3477

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68. C. G. Lawson
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- 71. M, I. Lundin
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R.C.

 
 

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