ORNL-CF-61-8-86.txt

 

UCN-2363
(3 11-60]

 

~ SEP 19 1961

MASTE!

 

OAK RIDGE NATIONAL LABORATORY )
Operated by

 

 

UNION CARBIDE NUCLEAR COMPANY o -'
. Division of Union Ccrl:ide Corporation 0 R N l
Post Office Box X - | CENTRAL FILES NUMBER

Oak Ridge, Tennessee 61-8-86

 

 

 

- | Externsl Trensmittel Authorized
DATE: August 18, 1961 - | | COPY NO. 5/,

SUBJECT: Thorium Breeder Reactor Evalustion. Part I. Fuel Yields and

_ Fuel Cycle Costs of a Two-Begion, Molten Salt Breeder Reactor
TO: Distribution
FROM:  W. L. Carter end L. G. Alexender

ABSTRACT

The MSER (1000 Mve station) is capsble of giving fuel wields of about T%/yx
(doubling time = 1k years) at & fuel cycle cost of approximately 1.5 mills/kvhr.
At fuel ylelds of 1 to 24/yr (DT = 100 to 50 years), the fuel cycle cost extrap-
olates to 0.65 mills/kwhr, et 44/yr (DT = 25 years), the fuel cycle cost is about

0.85 mills/kvhr, All systems vere optimized with respect to fuel cycle proc-
essing times.

| The effects on breeding perfbrmance of uncertainties in the epithermal value
of 1-233, uncertainty in velue of the resonance integrel of Pa-233, variable
thorium inventory in fertile stream and inclusion of ZrF¢ in reactor fuel were
eveluated. These effects may be sumerized as follows:

1. A +10% veriation in the epithermal value of n=233 from "recommended”

_ value causes & $2.5 to 13%/yr variation in fuel yield but only a +0.06 mills/kwhr
_ variation in fuel cycle cost.

2. Using 900'barns 1nstead of 1200 barns for Pa~233 resonence integral has-

only & small effect on breeding performance; the lower velue increases fuel yield

about 0.25%/yr and lowers fuel cycle cost ebout 0.0l milis/kwhr.

3. Doubling the thorium inventory adds about 1. Q%Jyr to fuel yield and
0.2 mills/kvhr to fuel cycle cost. |

4, TFive mole % ZrFq in LiF-BeFé-UFz fuel salt decreases fuel yield about

0. 5%/&r, but fuel cycle cost is negligibly affectea.*

TR

NonCE

- This tlocument ‘contains Infarmation of Q preliminary noture
and was prepared primarily for internal use at the Ock Ridge
National Lchoratory. It is subject to revision or correction
and therefore does not represent a final report. The information
is not to be obstracted, reprinted or otherwise given public
dissemination without the approval of the ORNL patenf brench,
Legal and Information Control Departmenf A

 
 

 

 

 

 

 

LEGAL NOTICE —

This report was prepared as an account of Government sponsored work. Nelther the United Stotes,

nor the Commission, nor any peraon ecting on beholf of the Commission:

A, Mokes ony warranty or representation, expressed or implied, with respect to the accurocy,
completeness, or vsefulness of the Information contained in this report, Ol'. that the use of
any information, - opparatus, method, or process dllelasod In this report may not lnfrlngt
privately owned rights; er c

B. Assumes any licbilities with respect to the use of, or lof damages resulting from iln use of

any Informetion, appurctus, method, or process di:eloud in this report. .
As vsed in the above, ““person octing on beholf of the Commission™ includes any omployoo or
contracter of the Commlulon, or employee of such contractor, te the extent that such omployn
or contractor, of the Commission, or employee of such contractor prepares, disseminates, or

provides access to, ony Information pursuont to his ompioymom or contract wlth the Commlulon, )

or Ml -mploymonl with such contractoer. -

 

 
 

‘-:"\ \S | :"J

1.0
2.0

3.0

CONTENTS
| Page |

Ahstract -nfl-----§ ------ Sao e e LT 1) D G i S Y D S G B G D S5 G S S NS a0 - o e @0 5w b 0 WP »l
Fbreward --‘--‘----F--------‘--------------fi--f-"-----,-- ------ momeewn 5
summary lv-----alwuof-fl--fi---anu-----------------------nfi,----m----------- 6

IntrOduCtion h.nnnnuun--,------n.--—----.-----------é-----—-n-i--q--- 8

Description Of;systém}-----q;q;---n—----;----------n-.u--una-u-.-----— 10

2.1

2.2
2.3

2.4

Ehy81cal systemqfi-wn-----auum--------—-—------—-p ------- - e o 10
Re&ctor COfe -------.-—---—-.--n--n---uunucn-------------------nn lo
Re&ctor Blank ---hu-fiu----------—--q--------nu-----—um-.- ----- - 12

ReaCtor coanOSition Vfi-----fl--“-----flfl------ ---------------- A 12

-Salt camPOSItion G S U W S et D U om0 b sk o D D D A D O B Y U 5 O ED S 0 BN G SN 0 S0 6 12

Chfimical Reprocessing.system T A DG O G O G S A T e O S A D B S S D U N 17
Fuel Salt Purification rrrrr T ry ey rr et vt ot Y L U L T D DL L 17
Fertile Stream Frocessing =e=weecremcescsccccccconesoseos —wmmme 1T

F18810n’éas R&mflval fi---------;----‘------ ----- T S T N A D D N N G R S 4 19

) Power Generation cycle (1 D P T O S R D G i G A T G D Sy D W G U S S o e Y.t 19

DeSign Bases &nd camputational.nethOdB ;---—-G-----;--------u-------- 20

3.1
3.2
3.3

3.4

3.5

3.6

| 331 |

3.9

Plent SiZe ceccwcccmwmccccwvsnmccncccsrenensccrescncnrsanescsnmemee 20
On-Site ProCeEEIiNg wemcmmmemmm oo oo o w oo o o -———————— we 20
Operating Conditions T L LT T - o
Product COmPOsition =e-eemmmmmomcmemmeccscemeaescnannean—- wnmaee 20

System Inventdry u-u-------f--nmnn------umhfi----nnu----—--q ----- 20

Neutron Losses -----u*fih----;-----u-h--p---------uu--------n--,—- 21
xenon-P0150ning -----d--,éé--n#haa---f-é--é—uuunn----u-u----d---- 21

Other Fission Product Polsoning ==eememseececcasmceosnnesnnonenem" 21

. Fhfil Proce531hg Losses duyn;—iw.d;--*u---------;-----h----fi-nff-- 22.

carrier and:nbderator Losses.;-------i--n---—--m-—----nu------ié— 22

Nuclear Data ---n----;----uin-n---miuunnaunu-fi-----m- ----- fi---q-- 22

Fuel Cerrier and_Blanket Carriér Cross Sections meeeneenemesemse. 22

.‘HuCIEar_Caléulatidns ---f------—é-—i-;,-------;----é-__---_.‘____ 23
Costs of Materials end Facilities end Interest Charges -==-=-==- - 24

. Fuel Stream'Proce881ng u;i-;--------in-u*ind-d--;n-’r-f-----.-..- 25

Fertile Stream Processing ~«ewecasccarececenanean cemmencemeeemnee 30

 
 

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' CONTENTS - contd

Reactor calc‘fl.ations --i-----‘-é-"flflfifl‘-fifl“---..--fl.:flrflh-.fl‘--fifll-‘;----.--.---fl‘
R 351 - € :

Reactor Size ._--‘_‘“J----“-:.--;_“k”-“_-”“.“"‘,-'-'-_-l.--‘;-"-'.-‘.,- -----------

) core Size : 'n--—-'-'fii--i-uéuoudo--'-—-uno;-u'um‘n-'--—----n--ii;-:- ----------

' manket Thicmess ---.'-----';;”‘.n--‘.-'_-‘--;,.-“‘-.,-..":-.‘t-”-".'-----------.-

B - Reflector Thicmess -i--lnt;nddh--'-a-éu;;néru;-.u-“—;nfi‘--—-Q--------—

2

-y

GMI calc‘fl_a,tions ----7->-“;-7'----—-'----;-—--'d;“-h-.---' -;é-b-‘-u‘;-a----fl

Inp]rt Data, “--‘.;fl---‘-r-.."l-r.-‘“‘-:..-;fl-‘--i-“fl;_.fl.;fi.fl..-l‘---.fl--lflfl-"-.flu-fln

. mtput Data, ---nfl--u--------uu---u—----------ufl--flwfl--w-m---- ------
"Cornpone Unit Cell Calculation.---------..----.------- ...........

’ Inp'ut Data -—-qn----------fn-u-h--------nfln----------uu -;--.------

output Data, -------Q------.--f----------;;;----------- ----_--fld-fl-hfi-;

‘‘Reaction Rate Coefficients -------;-;;----;--;;;-----;-a---------

4.5

5.0

’ ou,tput mta.'-fi---------nu--nqd—-'-:‘-fi-.i---é--dh'-'i;----‘i-;-_--j-‘--h--'-r-‘

Fuel

'?-5;1

6.0

5.2
23
5.k
23

5-6

5.7

Parameter Studies and Results ----------;--;---;---------.-;-;--------’

6 1

Equilibrium Reactor Calculations (ERC-5) Ceesiudmumem e anaa—e

) Inpuvt Dat& --------------an--"'---'----6------"-"-'--'-—-.--}--n-'-; ------ -

Solution of Poison Fraction Equation weeeemceccececescecnccmcones
Fission Products Included in Poison Fraction Calculétions -------
Gas Spéfging and Effective Yield evecocurmacmcccrcinncmcncccccan"
Fission Products as 1/v Absorbere -=esmmememroe——-— ————— o
Fission Rro&uct Resonance Absorptions Included 1n Poison S e e

Fraction Celculetions =eesmecsccocmmeccnacnmenanunnnssnnanennesen

-Use Of Figures 5 1 and 5 2 - o S S8 . ;?-----;-----------.-“fl.-;--flufl

“Results of Equilibrium Reactor Calculations - 1 o 0 0 1 1 0 o e 0

) _60101 F\l&l cyc1e Times ------------------‘------.----------. ----- -

6.1.2 HEutron Balance -------------------------;;---------.--;--
Resonance Absorption Cases e 0 0 0

 1/v ADSOTPLion CASES mmmmmemcmcscemmecsessmononnnsmn—-————
6.1.3 System Inventory ----sseececcac=a- - 20 0t e o 0 0 e
6.1.4 Fuel Cycle COSt me=mmmmememmmen ———————————————— crmnemnmn—-

Page
35
35

35

35
35
36
36
36
36
39
%0
4o
40
41
k1

Stream Poison fiaCtion Cal'cul&tions il'0fi"fl----'-‘---'u------'l--:-‘-'-———---'

POison Er&ction n;a;'.;-ui-ii—nufin—---n--u-n-uu--n.n-.u--_---_--_;;---

43
45
47
4T

%%&fl&fls

29
60

62

 

 
 

6.2 Pbison Eraction Studies in which.Fission Product

. Resonance Absarptims a,re Inca'uded --“nn--an---a---c--u—-----—,

Fuel Yield Vérsus Pbison Eraction o et 0 1 0 o
Fuel Salt Discard Time as & Function of Fuel |
Cycle Cost e e e 0 e
FR :6 2.2 Econamic PErformance e e e e e e
5=6,3;,Pbison Fraction Studies in ‘which Fission Products were |
o Considered to be l/? Absorbers e e e e o 0 0

6 301 ECOIIOIIlic Performance flfi-’a#w---mh-----—------nn—------uuunn

w_,6rh,'EffEct on Reactor Performance of Varying Thorium.InVentory ——

. 6.5 Effect of Valufi of n~233 on MSER Performance ---,-.--_----;----_-

6. 6, Effect of Value of Pa-233 Resonance Integral on R
- - MSER PErfbrmance 00 .00 . o SO o o 00
6.7 ‘Effect of MSER PErformance of Adding Zth to
 Fuel SELE mrmmmcasesmte e ————————————————————————————
7.0 Conclusions_-—---e-qu--_.-g----,-,.-,-—---,-;.-..--,fl-.---.,.;,-.-..
8.0 References ammemmsmmmmemmmacmmceceesseeseeennee.——"———————————————————

9.0_ Append_ix -u‘mm‘_u-m-u-u?wQflfln‘punp‘”nu---_------_-.----q--—-nn-------.-'-_---'--n-._q

- Page

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83

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| Breeder Reector Program, en eve.luation of types of reactors cs.pable of ef=
_'ficient utiliza.tion of thorium was initiated at OBNL in- July 1959 Included
in this eveluetion were studies on the Aqueous Homogeneous Ereeder Ree.ctor
.(ABR) , Molten Salt Breeder Rea.ctor (MSBR) ’ Grephite-Modereted Ges-Cooled .'
" Reactor (GGBR ) Deuterium-Modereted Ge.s-Cooled Breeder Res.ctor (DGBB) and |
| Canedien-Deuterium-Ura.nium Reactor (CANHJ ).

EHORD

As pert of the ORNL responsibility for guiding the AEC Therma.l Thorium

 

This report presents the results of the MSER evaluation. ' A comparison
of ell five of these reactors ha.s been presented 1n two previous reports by
this study group. The reader is referred to these reports for en s.ppreci.-
etion of the performa.nce or these several systems These reports are' .

L. ‘G.‘Alexslrider," et al., Thoriun Breeder Reactor Evaluation. Part I.

Fuel Yields and Fuel Cycle Coste for Five Thermal Breeders, ORNL=CF=-

-3-9, March 1, 1961.__ - : , o

L. G. Alexander, et al., Thorium Breeder Resctor Evelus.tion. Part I.
Fuel Yields end Fuel Cycle Costs in Five Thermal Breeders , ORNL-C‘F-
61-3-9 1Appendices, Pert I), Ma.rch 1., 1961.
 

 

A two-region, molten sa.lt breeder reactor (MSBR) ha.ving core dimensions approxi- .

mately T.7T £t diameter ’oy 7.7 ft high and surrounded on the ends end sides by & 3-ft-
thick.blanket was studied for determination of its breeding perfbrmance and fuel cycle
cost. ‘l‘he oore composition uas a.pproximately 16 vol % mel-bea.ring se.lt, 6. T vol 4
fertile stream ‘and 77 3 vol % gra.;phite; side blanket composition was 90 vol 4 fertile
strean amd 10 vol % graphite. Basic criterie. of the study were that the reactor com-
plex be cepable of producing power et & ra.te of 1000 Mwe and the.t chemica.l processing
be ce.rried out on site. Two reactore were required, producing stee.m at 1800 psia. a.nd
1050°F.

The fuel salt passed through the core a.nd upper end 'ble.nket in some 90 two-pass,
bayonet tubes made oi‘ impermeable graphite which e.re inserted in openinge in the
graphite moderator. The region between the core and reoctor vessel end the annuli
between the fuel tubes ‘and. moderator are filled with fertile material. To minimize

dnventory the ruel stream pump and hea.t exchanger a.re mounted directly a.borve the |

reactor core.’ ST S : - : . . : o
The fuel salt was & 63 -37 mole % mixture of LiF-BeFa containing at equilibrium
ebout 25 gm U per kg salt, of which ebout 18 gm was U-233 and the remainder was higher
ieotopee. The fertile salt was & 67-18-—15 mole % mixture of LiF-BeFe-ThF;. At equi-
librium the fertile stream contained from T70 to 2400 gm U-233 plus U-235 per tomne
salt. The fuel salt uas processed for fission product removed by the fluoride vola-
tility process and the HF dissolution process. A portion of the fuel salt was dise

carded during each ;proceesing cycle for removal of fission products not removed ‘by

HF dissolution. The fertile stream vas processed by fluoride volatility only,

fission product accumilation in the fertile stream was maintained at & tolerable

level by discarding the fertile galt inventory. on & 20-year cycle. In this reactor

only 1.3 - 6. 6% of the fissions occurred in the fertile stream. |
Nuclear calculations were performed using the 34-group, multiregion GNU progro.m8

for the IBM-O4 and the Cornpone program’ for the ORACLE. After sttaining eriticality

in these calculations s Turther computations were made using the ERC-5 programlo for

the IEM-"{OM to determine the equili'brium condition. ‘It is the equilibrium results

- that are reported here.

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The MSER is capa‘ole or breeding over a wide ra.nge ‘'of operating conditions
giving fuel ylelds as high as ebout 7%/year for & dou‘bling time of ebout n.5
full-power years. At this high yield, however, a prem:l.um fuel cycle cost o:t' ep-
proximately 1.5 mills/kwhr is incurred principally beceuse of high fuel stream
processing 'char_ges; The fuel éycie cost was__ optimized by determining for each -
fuel yleld the most economic combination of fuel stream processing cycle time and -
fuel salt discard cycle time. The fuel yleld: was made to vary by essuming several -
veluef of the fuel stream poison frection and the fertile stream cycle time.

“In the realm of more economicel operetion, fuel cycle costs as low as 0.65

" mms/kwhr ere predicted at fuel ylelds of 1 to 2%/year. ‘When the fuel yield is

h%/year, the fuel cycle cost is approximately 0.85 mills/kwhr. At this latter
condition, ‘the income from sale of fertile materiel Just offsets the snmusl in-
ventory charge. : T ST -
Ca.lcula.tions for -a represente.tive set of" opera.ting conditions were made to.
evaluate ‘MSER performance in the 1:lght ‘of uncertainties in nuclear data (value of
7=233 end the resonance mteg_'e.l ‘of Pa-233), variable thorium inventory end ed-
dition of ZrF4 as & -sfabiliz:l.nig agent for the reactor fuel. Eta iralues at epi-
thermal energies within 110% of the values ‘recommended for this study were employed
in nuclear ca.lculationsgiving'a.tz.5 to"":t3%/year verietion in fubl yield; corre-
sponding fuel cycle costs were negligibly affected (+0.06 rmills/kwlhir-). Reactor
performance using & resonance integral of 1200 barns for Pa-233, used for this
study, was compered with that for'a 900-barn value; fuel yield was improved asbout
0.25%/year with a negligible lowering of the fuel eyele cost. A lower thorium
:anentory (140 tonnes vs 270 ton.nes) decreased the fuel yield ebout 2§/yeer with

. a corresponding decrease of 0.2 mills/kwhr in fuel cycle cost. A representative

calcula.tion in which 5 mole % ZrFq was added to the fuel selt mdicated that the
fuel yleld would be lowered by ebout 0.5%/year and that the fuel cycle cost would
be negligibly affected as compared to & simllar case containing no zirconium.

 

*Based on a pla.nt factor of 0.8

 
 

1.0 mTRowcTIoN [
The vork on the Mblten Salt Breeder Reactor (MSBR) reported in this memo-
randum is a portien of ‘& more camplete study on thermsl breeder reactors, which
includes the Aqueaus Hbmogeneaus Breeder Reactor (AHBR), the Liqnid Bismuth o

Breeder Reactor (LBER), the Gas-Cooled Graphite—Mbderated Breeder Reactor (GGER), .

‘end the Deuterium-Moderated Gas-Cooled Breeder Reactor (DGER). . The important -
results of the complete study on all-five'reactors is reperted in ORNL CF~61-3-9'
by Alexand.erl et al; 1t is the purpose of this: memorandnm to present more detalled
date &and calculations on the MSER than those included in the reference memorandum.
It is adviseble for the reader to examine ORNL CF-GL-3-9 in_conJunntion‘with this
memorandum in order to meke & comparison of the several thermal breeders end to
‘obtein informstion on the MSER thet may not be repeated herein, |

The MSER was examined with. the viewpoint of obtalning a relationship between

breeding potential and economic performance. . Breeding potential is related directly

to neutron economy and is therefore associated with the composition and deSign of

the reactor. Economic performance is determined by the annnnlrcherge~on;such~

~ items as the capitel 1nvestment in the'reactor'inetallation; cepitel investment.

in chemical processing plants, operation of these plante, inventory of valueble

materidls-(esg., uranium, thorium, fuel carrier,selt and fertile carrier salt),

use of thecse materials, end waste disposal. On the other hand, income from bred,

fissionable material in excess of that required to refuel the reactor 1is credited

to the economic performance. Two of the ebove charges have not been included in

this cost analysis-because no reliable cost data are available; these are the

capital investment in the reactor installatian -and waste ‘disposal charges. In.

defense of omitting waste disposal charges, it might be said that'since all

westes are solids the disposal charges will be & very emall’ fraction of the

total charges. It 1s observed that the remaining cherges are concerned with the

reactor fuel cyele_and henceferth are refErred,to as fuel cycle coste. .
In order to make a breeding system of. the MSER, it is necessary to. exerciee .

control over those neutron poisons that are smeneble to control; some poisons,

such as reactor structural materials;, asre fixed by design reqnirements. A

eignificant advantege in neutron economy is realized by controlling poisoning

from fiSsioniprodncts by:ehemically-prOcessing fuel and fertile streams for

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their removal. It is epparent that the system in equilibrium may be operated
et any desired poison level.between thet corresponding to some practical minimmm
end that of complete burnout of fission products. It is customary to identify
fission product poison level in & reactor &s the poison fraction, which is de-
fined es the ratio of neutrons ebsorbed in fission products to neutrons ab-
sorbed in fuel. - ”

There 1 en inverse relationship of poison fraction to breeding and economic
performance. In order to maintain high breeding performance, it is necessary to "

-chemically process fuel and fErtile streams on e relatively frequent schedule at

the expense of high fuel cycle cost. On the other hsnd, less freqnent processing |
lowers the fuel cycle costs but has an adverse effect on breeding performance. |
The fuel cycle cost assocleted with each poison fraction can be optimized by the
proper choice of fuel stream cycle time and fuel salt discard time. (See Section
2.3 for & discussion ‘of the chemical processing system.) In this study’all fuel

'cycle costs have been optimized with respect to fuel stream processing conditions

but not with respect to fertile stream processing conditions. The fertile stresm
conditions were included as a'parameter-study'in vhich & series of fertile stream
cycle timee in the renge 35-200 days vere studied for each velue of fuel stream
poison fraction in the" range 0.011 - O. 065._ ‘The pertinent results are exhibited
as plots of fuelecycle cost (mills/knhr) versus fuel yield (%/yesr) end poison
fraction. |

 
 

S2a @ysicai System S | _ o

The molten salt breeder reactor examined in this study is based upon the
design of thPhers nt3 and 1s pictured schemetically in Fig. 2.1. The reactor
is cylindrical with ) core T.66 £t in diameter -and T 66 £t high. The core is
surrounded on the sides and ends hy & 3-ft-thick'blanket. A.l-ft-thick graphite
reflector surrounds the hlenket on the sides and the ende The reactor, heet
exchanger and circuleting pump are arranged in a compact, vertical configuretion
to mdnimize the fuel velume.. Surge volume fbr the system is provided 1n the o
chamber housing the pump impeller.,_'- o o

| Reector Core. The reactor core is made entirely of graphite formed by
‘essembling 8«in, square prisms. The corners of adjacent prisms are machined to
form verticel passages of circular cross section ebout 5 in.rin diameter, - The.
fuel salt passes through the core in tubes of bayonet construction which are
1nserted into these machined vertical passages;_thesiuel\tubesgarevmade of im-
permesble graphite. The outer tuhes'(see Fig. 2.1) have inside_dianefers,of' ‘
3.75 in. end walls 0.75 in. thick. They are joined to an INOR-8 metal hesder
by means of flanges, frozen-plug seals, brazing, or transition welds.f_Thesegwefl
Joints are presumed to be suhstantially leakproof. The inner tubes have inside
dlemeters of 2. in, apd walls 0.25 in. thick. They are joined to the inner
plenum of the metel header by slifi Joints; these joints need not berleakproof
since some bypass leakage et this point can be tolerated.. The reactor centains'
| approximately 90 bayonet tubes. | o |
' Sufficient clearance between the fuel tubes and graphite moderator is
provided to allow for differential expansion between the moderator and the ;
metellic fuel plenum. Fuel salt enters at 1125°F, passes down through the
ennulus in the bayonet tube, rises through the inner tube at 20 ft/sec, and
exits at 1300°F. It is collected in the plenum end passes up through e duct
to the impeller of the pump from which it is fbrced through the tubes of the
~ heat exchenger. After leaving the hesat exchanger, the cycle for the salt is
repeated. The ealt circulates at approximately 50,000 gpm, removing 1070 Mw‘

of hesat. The heat exchanger conteins approximately 8100 tubes (INGR-B) vhich ; e

are 0. .375 in. in outside diemeter and have O. 028 in. walls. The shell side of -/
the heat exchenger contains molten sodium. ' o S

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FUEL SALT

(PRIMARY COOLANT) * I | ._'gRAPHITE-METAL
| o - M - JOINT .

© GRAPHITE TUBES ma

VESSEL ——>]

_ | . UNCLASSIFIED
- 11 - - ORNL-LR-DWG. 46040R

~ SECONDARY .
COOLANT_

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| Ficj. 2.4. Molten Sait Breeder Reactor.
 

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Reactor Blanket. The ma,jor portion of the fertile salt circulates through
the side and end blenkets 3 however, 8 smsll portion bypasses through the core

 

- in the passages between the fuel tubes and the grephite moderator. In its

passage through the reactor the fertile salt temperature rises from 1150°F to

. 1300°F; this sensible heat is then removed in e sod.ium-cooled heat exchanger.

The salt circulates et s.pproximetely 3900 gm end. removes about 112 Mw of heat.
This 1s about 10% of the tota.l res.ctor energy; however, only gbout 1.3 =~ 6.6% of

the reactor energy originates from :E‘issions in the fertile stream. The hesat

exchanger contains approximately 1000 tubes (INOR~3) which are 0.375 in. in

- diameter and have 0 028 in. valls. |

\

Reactor Composition. The s.pproximate volumetric composition of the reactor

- core is as follows: 16% fuel stream, 6. 7% fertile stream, and T7. 3% grsphite.
VThe volumetric composition of the side blanket is 90% fertile stream and 10%

graphite. The top end blanket contains both fuel and fertile stream; the volu-

metric composition is 16% fuel stream, Ti% fertile stream, and 10% graphite.
Additional date on the reactor and heat removel system are given in

Table 2.1. |

2.2 Salt Composition

The fuel salt consists of & mixtnre of 63 mole % LiF end 37 mole % BeF,

| containing sufficient UFI; (equilibrium mixture of urenium isotopes) to. make the

system critical - a.bout 0. 35 mole % , -
The fertile stream has s. ‘besic composition of 67-18-15 mole % LiF-BeF -ThFh

| The equilibrium mixture of course contains Pa-233 ’ uranium isotopes snd & small

concentration of fission products. - The urenium content of the fertile stream is
mainteined at e. quite low level by the efficient fluoride vola.tility processing

method (see belov) ; therefore, it is not extremely important that the fertile-

~ stream volume be kept low. In fact, in some ceses it is desirable to have a

 large excess fertile-stream volume to decrea.se neutron losses by protectiniun
cepture through the dilution effect. '

| The distribution of fuel- and . fertile-stream volumes inside and outside the '

MSBR 1s tebulated in Ta.ble 2.2

<)
oy

wd

al)

 

-13 -

Table 2.1. Molten Salt Breeder Reactor Flant pata(®)

General

Station electrical powver, MwE |
Station net thermodynemic efficiqncy; g

 Number of reactors per station -

' Thermal power per stetion, MwT
" Fraction of electrical power fed back into plant

Fuel

‘Mean heat capacity, Btu/lb-‘F
-Power density in portion of fuel stream

Geometry of core

Moderator L

Volume frection of moderetor in core
Diameter of ¢ore, ft

Length of core, ft

Thickness of blanket, £t

Volume fraction of'moderator 4in side blanket

Volume fraction of moderator in end blanket
Reactor vessel materiel e
Reactor vessel thickness, in. = =
Mean pressure in reactor, psia '
Diameter of core fuel channels, in.

Stream

Fuel‘carrier o

Density (1200°F), /et

Fraction of thermsl power removed by
fuel stream heat exchanger :

A

external to reactor, Emt/ftB

,Liq_uidus temperature, (b) °F

- Stetion flow rate,- ft3/éec |

Velocity (ft/sec) of fuel stream in
Core .
 End blenket

_11§;5

1000

42.3

P

1182

) :0003 :
~ eylinder (L/D = 1)
~ graphite

T.66

. T.66

0.10 ’
0010 o

- INOR-8

1.375

< 100

3.75

63 mole % LiF

37 mole %_EeFé

o091
7 ':‘ 005""""

7.6

178

20
20

 
 

 

 

ik

Table 2.1.  Contimued .
Heat exchanger data: .- ;
 Tube outside diemeter, in.
“Tube wali'thickness, in.
 Material . o
T&be velocity, ft/sec '
Flow rate, 1b/hr
Fluid temperature in, °F
Fluid temperature out, °F
Pressure drop, psi
" No. tubes per exchanger
Tube length, ft |
Tube bundle diameter, in.
Inside film coefficient, Btu /hr-ft2-°F
Tube wall coefficient, Btu/hr-ft °F
 Scale coefficient, Btu/hr-ft -°F
Outside £ilm coefficient, Btu/hr-ft2-°F
Over-all coefficient, Btu/hr-ft2-°F
Outside tube sarea, ft2

Fertile Stream

Fertile stream carrier

Density (L200°F), 1b/£t3
Mean heat capacity, Btu/lb-°F

Fraction of thermal power removed by
fertile streem heat exchanger

Fractlon of fission power produced in fertile stream
Liquidus temperature, (b) °F |
 Stetion flow rate, f£t3/sec
HBeat exchanger data: .
- Tube outside diameter, in.
‘Tube wall thickness, in.
" Materisl
~ Tube velocity, ft/sec

2620

tube. side shell side

0.3715

INOR-8 - -InoR-a

 }leJ u$x1J
1300 - . 900 -

125 1175
7% 100

810

11.13

69

8020

7080
110,000

48,900

8320

67 mole % LiF -
18 mole % BeFp -
15 mole % ThF)

. 192
10.32

009

932
22.6

tube side - shell side
0.375 o
0.028 o
INOR-8 «  INOR-B
1k.1

O

<)
=}

4

o))

o

-15 =

- Table 2.1. Contimued

tube side'

 

shell side

Flow rate, 1b/hr o 598 x 10° 4.48 x 10°
‘Fluid tempeia.ture, in, °F - : 1300 900 |
Fluid temperature out, °F | 1150 1175
Pressure drop, psi ' ' 109 100
No. tubes per exchenger - 1050

 Length of tubes, ft | - 19.7
Tube bundle diemeter, in. o 27

~ Inside film coefficient, Btu/hr-ft F 5550

. Tube wall coefficient, Btu/hreft®-°F | 5660
Scele coefficlent, Btu/hr-ft2-°F 10,000

- Outside film coefficient ’ Btu/hr-fta-’F - 40,000
Over-gll coefficient, Btu/hr-fta-'F . - 1845

| Outeide tube area, fba | 1925

 

(e) A number of items in this tabulation are from a study by Spiewak a.nd

Parsly. 1k
(b) Temperature at which LiF precipitates.
 

- 16 -

Table 2.2. Distribution of Fuel- end Fertile-Stream Volumes
in thé,Mplten Salt'Breeder Réactor |

~ Volume - ‘Volume per 3

',"'_ '",: PR |  fraction . .. station gft )

Fuel stream.in
 Core o ';-,; | _ 016 | 13
Upper end blenket | 016 884
Lower end blanket .0 ‘-,¢_1 o |
" External to'reactdr  | _  '.'f ";; ;._ 286,6 j .
' Dumpftanks and_miscellaneous } | | '»i _;_i,ha;a -
Total | - S 53002

Fbrtile stream in |

Core - % ~0.06T 95

Upper end blanket | 0.75 | { 409
~ Lower end blanket 0.90 |

Side blanket - 0.90 - 2470
' External to reactor | | 3026

Totel - | a 6000

o
 

w)

)y

wl)

@ lT -

2.3 chemical Reprocessing System

_ A flow diegram.of the chemical reprocessing system.is shown in Fig. 2. 2.
The processing operation consists of three parts: fuel salt purification,
uranium recovery from the fertile stream, and helium sperging to remove fission
geses from the fuel salt. ’

,flr Fuel Salt Purification. The fuel salt is purified in the fluoride voletility-
HF dissolution process by pumping a side-stream of the circuleting molten salt
through the processing plent in & specified cycle time. The cycle time is &
function of the poison fraction et vhich the reactor is permitted to operate,
which in this investigation is & paremeter. o

~ The first step in purificetion is to fluorinate the molten selt with ele-
mental fluorine to voletilize UFB This urenium.hexefluoride is then burned in
hydrogen to produce Urh, which is recycled to the reactor after dissolution in
the recovered carrier salt. Uranium-free salt, conteining fission products,
flows from the'fluorinstor to the HF.dissolution.step.s Here evseperetion is
made between the salt end the bulk of the fission products. The carrier salt
is dissalved in a 90% EF-10% H,0 solution leaving fission products, principally
rere earths, as insoluble material. The carrier salt is recrystellized, fortified
vith recovered Urh, and recycled to the reactor. In order to purge those fission
products which are not removed 4n the HF dissolution step, portions of the fuel
salt ere periodicelly removed end fresh meke=up salt ie added. The fission
products purged in this manner Anclude msinly the elkali metals end elkaline
earths such as Cs, Rb, Sr, Ba, ‘I'e, Se, Nb, Cd, Ag, Te, ete.. o

The fuel salt replacement cycle time depends upon the fuel stream cycle
time end the poison fraction. It is possible to achieve &’ specified poison
fraction with severel combinations of fuel stream cycle time end fuel ‘salt re-
plecement cycle time as is shown in Figs. 5.1 and 5. 2. The prqper replacement
cycle is determined by optimizing the fuel cycle cost with respect to several
combinations of the twvo cycle times. T

 

Fertile Stream Processing. The fertile stream is processed in the fluoride

1soletility‘step only. The salt is circulated at & specifiedcrete through a
fluorinator where contact with fluorine gas volatilizes UFg. The sslt'then
returns directly to the blanket without additional treatment. The UFE from the

fluorinetor is reduced with Hé to UFh which 1is blended with UFh recovered from

the fUEl selt for recycle to the reactor. Excess production 1is sold.
 HF
' To Recovery

| ,UF.-',-UE.-.

REDUCTION

 

 

 

UF,
ety

FLUORINATOR
| [ Lie-Ber-mE,
- UF,,FPs,Pa |
. LiF-BeF-ThE,
' Make-up

 

 

 

 

 

 

 

 

 

 

 

Solf Dlscurct To
- Remove FP's

 

 

(20year cycle)

 

 

 

 

 

 

UF4H_

 

FIG.2.2 |

 

CFUEL
MAKE-UP

 

'-U.F'afl

UNCLASSIFIED
'ORNL~LR-DWG 54768

90% HF

 

 

 

 

  
  
   

 

100

 

 

 

 

 

 

 

 

 

Excess |

" Production to | ,

. Sales -

 

 

 

salt+ —1 -}

[_

' EVAPORATOR

 

 

 

 

 

 

 

 

 

 

 

 

 

 

I

' 'CONDENSER

 

“to Wuste’mu’ { \ o .
{ )lny,i » —HE.

 

 

-gr~

 

 

tare earths) -
Do STORAGE

 

 

 

 

 

 

 

 

 

T LiF-BeFe | )
Salt Storage | /

 

 

 

. LiF-BeFe
. make-up. .

Salt Discard
to Remove Soluble -
Fissio_n Produc_ts

SCHEMATIC FLOW DIAGRAM OF MOLTEN SALT BREEDER REACTOR FUEL& FERTILE STREAM N

PROCESSING

C)

 
)

)

“19 -

~ In the MSER at- fertilé stréa.in cycle times less than 100 deys such & small
fraction of fissions (<6.6%) occurs in the 'f‘ertilefl stream that it is not neces-
sary to purify the selt in e HF dissolution step. The fission-product build-up

is slow enou.gh that the:l.r level cen be comreniently controlled by replacing the

salt on & rela:hively long cycle. A 20-year cycle has been specified in this

~ study.,

It w'n_l 'be observed that protactinium is net removed from the ferb:l.le selt
in this process. FProtactinium builds up in the _salt until its decay rate is

| Just equal to the U-233 production rate. The effect of Pa=233 on the neutron
. economy is controlled by a.d,just:lng the volume of the fertile strea.m, la.rger

volumes g:lv:lng fewer neutroan losses ‘to protactinium.

. Fission Gas Removal, Fission gases ere removed from the fuel and fertile
streams by sparging with h‘elium;_ “Xenon, krypton and the halogens are expected
to be removed in this way, The io;_ff—ga.s is pa_ssed through charcoal beds where

~the fission gases are sbsorbed, Helium ie recovered for reuse,

2.4 Power Genera.tion Cycle :
 The steam cycle of the TVA John Sevier power plant vas used es a model for

the MSBR concept. __5_ Steam canditions ere teken as 1800 psia. end 1050‘F, the

condenser pressure is 1.5 in, Hg Additiona.l data on power generation equipment
are given by Splewak and Pa.rt:'»ly.:"lL
 

- 20 -

.

. 3.0 DESIGN BASES AND,CGMPUTATIONAL'METHODS_,_-_
3.1 Plant Size - |
‘Based on & study by Ro‘bertson,2 it was assumed that future power stations’
in the United States would heve & capacity of the order3of_1000"MweQ"Consequently,
- this size was chosen for this study. Also it was assumed that anyone building'e

‘plant of this size would be unwilling to insta.n the entire Toad n & eingle re-
actor; therefore, at least two reactors are specified for each station."" i

3. 2 On-Site Processing

Qn-site chemical reprocessing vas chosen for the station. This method lends
itself to better control and. definition of in-process inventory. rReaSOnebly'
reliable .cost estime.tes11 are aveilsble on fluoride volatility plants for proc-

| essing core and fertile streams..
3.3 gpereting Conditions

All calculations vere made for continuous, steedy-state operation of the - a

reactor complex. To avoid complicated calculations of startup and shutdown, it 7
was assumed that the resctors would‘be continuously fueled end processed, and - w

that the operation had been going on sufficiently 1ong for all fission products
and hee.vy isotopes to be in equilibrium, ” - |

3.4 Product Composition

The product composition may vary between the limits of almost pure U=-233 to
spent fuel.- However, in a many-reactor system complex, the fuel yleld (or doubling
time) is unambiguously defined only when the product hes the same composition as
the average composition of the entire system; i.e.,-reector plus'chemical proc-
essing systems. Calculated portions of the recovered spent fuel end of the bred
- meaterlal are removed as product et the UF5'UEu reducticn step. The product is
an equilibrium.mixture of uranium isotopes; viz., U-233, U-23h U-235, and U-236.

3 5 System Inventory

w

| ' In order to be consistent and unambiguous in the definition of fuel yield,
the inventory should include all fissionable and potentially fissionable atoms
'(05233, U¥235,'and Pa-233) in the entire system., Included in the inventOry is
any fuel tnat is reserved to allow reactor operation during shutdown of the
chemical processing plent. In this study & 30-dsy fuel reserve was chosen.

<€)
 

agl)

wd)

3.6 NEutron Losses

: Fission-product poisoning in this reactor was bssed on a study'by ‘Burch, -
Campbell, end Weeren,3 vho msde & study of the cumulative effect of k4 isotopes
divided into four ‘groups. This phenomenon is.discussed more completely in
Section 5.0. | | T |

Xenon Poisoning. It was assumed that xenon could be contimuously removed
from the circulating fuel byigssfsparging sndfimsintsincd:st*a.level‘snch'thst:
the neutron loss to xenon is O. 005 neutrons per fuel absorption. Since there
ere so fev fissions (<6. 6%) in the Pertile stream of the MSER, 81l xenon losses
were assigned to the fuel ‘stream. o Lo - '

. In choosing & vslue for xenon losses, it was assumed that (a) neither Xenon
nor iodine is ebsorbed by the grsphite moderator or otherwise collects at the
interface between salt end graphite phases; or (b) if the pores and vscancies -~
in the graphite sre'sccessihle to Xenon and iodine, the rates at which they
diffuse into the pores are very much slower than the rates at which they ere
stripped from the circulating strcam.by‘the sparge gas.' -

Other Fission Product Poisoning. Concerning the effect of the hh fission-; i
product isotopes studied by Campbell, Burch, and Weeren,3 four groups were dis-
cerned end treated separstely.l The first comprised noble metals which vere

‘assnmed to plete out on the cold zones of the circulsting system; the second

comprised halogens which were assnmed to voletilize dnring the fluorinstion

step._ A.third group which is solnhle in HF snd therefore ‘not removed in the
diesolution step comprised the slknli metals (notebly'Rb cnd Cs), the ‘elkaline
earths (sr, Ba, etc.) and a8 miscellaneous group (Te, Se, Nb, Cd, Ag, Te, etc. ).
This soluhle group is removed by replscement cf the fuel salt on some specifiedi'_

,cycle.r The finnl group is the rare earths which sre removed by precipitstion
'_during the HF disscdution step. The poison frection is thus not & simple function
~of the processing rste snd is computed as described in Section 5.0 helow. |

Fission products in the fErtile stream are. controlled entirely'by the

20-yesr throwsway cycle of the thorium carrier since this stresm is not chemi-
cally processed for fission product removal.Y“__“_‘_s__:_
 

-Fuel Processing Losses. Fuel processing losses'are based upon laboratory
‘and pilot plant data, which have indicated essentially quantitative removal of
uranium.from the salt by the fluoride volatility process. Consequently, losses
that occur will ve almost entirely in the UFE-;—UFL reduction step. It is be-
lieved that on large-scale operation these losses can be made quite small -—
of the crder of 0. 01% of throughput. | '

Carrier and Moderator Losses. Neutron.losses'to the carrier salts are
based upon the use of a feed salt in which the 1ithium component is present as.
99.99 at. % L17 -and O. oL at. % Lis. ‘A salt having & lower L16 concentration

 

_would be desirable, however, it is questionable whether or not the premium,price

for such a selt is Jjustified by the increased neutron economy

In this study the mono-energetic capture cross section of grephite was taken
to be h 2 mb at’ 0. 025 ev end was assumed to vary inrersely with the velocity of

the neutrons.‘
3.7 Nuclear Data

Nuclear cross sections for this study were compiled by Nestor.u ,Fbllowing'

the recommendations of Fluharty and Evans,5 & value of 2. 28 was assigned to eta,‘

of U-233 et thermal energies. The resonance integral of Pa-233 was assumed to
be 1200 ‘barns, Allowance for resonance saturation (self-shielding) was made |
only in the case of thorium, end here Doppler broadening was algso taken into ,
account. Epithermal cross sections of other isotopes of interest were adJusted
to egree with ‘the resonance integrals tebulated by Stoughton and Balperin.6 |
The composite 2200-meter cross section of fission products (exclusive of Xe,
Smélfil, end Sm-lhg) was taken a8 50 “barns per fission and assigned to an -
artificial element celled “fissium."” A.resonance integral of 170 barns per '
fission was assigned to fissium‘as suggested by'the work of Eephew;T_
in the. computation of poison fraction, resonance integrals had to be assigned
to each individual fission product. Available values were taken frcm Nephew,
1f no value was reported, it was calculated from available data.

However,

Fuel Carrier and Blanket Carrier Cross Sections. The fuel carrier, which
is composed of a mixture of Li7 Li6 Be, and F atoms, is treated as & single,
pseudo fuelesalt etom in the nuclear calculations.: It is convenient to do this

because the GNU code is limited in the.number of elements for which ebsorptions

G

»

<)
 

w)

wl

)

..2‘3 .'.-

cen be celculated.;_conzj:',equently, lunping these elements saved space on the' tape
for needed calculations. A pseudo cross section i’or the fuel salt was obtained
by normalizing the cross section to the besis of one atom of L17 eand summing the

results. In the norms.lization the cross section of each atom was multiplied. by

the atomic ratio of that particular e.tom to Li7 Tne atomic concentration of -
the selt is then expressed es the atomic concentration of I..:L7 in the salt. The
lithium component of the sa.lt waE assumed to be 99 99 e,t. % 11l ’

The rertilef-stream cerrier was trea.ted in e similar manner with the Cross -
sections of each component atom normalized to the basis of'an' etom of thorium.
The fertile~stream carrier contains ‘I.iT Lié,' Be, F, and Th atoms. The atomic
concentration of the carrier is then expressed as the a.tomic concentration of
thorium in the salt. The :l..i7 purity 1s the same &8s used in the fuel carrier,

3 8 Nuclea.r Cslcula.tions

Nuclear cslculations on the MSER vere performed. using two different reactor
codes: the 31+-group GNU code8 for the IBM»TO& and the COrnpone code9 for the
ORACLE. The use of the two codes expedited the calculations. The reactor was
first treated in sphericel geometry as a homogenized system using GNU, end &
criticality search was made to determine' the 'criticsl cOncentretion of thorium
and protectinifim in the core. The dia.meter of the equivalent sphere was te.ken

| . &8 1,09 times the cylinder diameter. This informa.tion was then nsed in Cornpone

celculations, &lso in spherical geometry, to determine the thorium concentration
in the core of the critical heterogeneous reactor. The heterogenelty of the
MSER could be studied on Cornpone through the use of "diss.dventege factors";

- disadvantage factors could not be a.pplied to GNU. Since all pa.rameter studies

wvere to be made on the equilibrium reactor s the eritical res.ctor concentre.tions :

from the cOrnpone ca.lculation were-used s.s input information for an equilibri\m

reactor calculation using the ERC-5 code ~for the. IBM=704. This calculs.tion :
determined the concentrations snd elemental neutron absorptions in. the critical, iy

' equilibrium reactor .

A more deta:L'I.ed discussion of the nuclear celculstions appea.rs in Section 3.2
of Ref. 1, . | _ |

The disadvanta.ge factors mentioned above vere used to relate the concen- 1
trations of the homogenized reector to those of the heterogeneous reactor. These
fectors were determined in e lattice=cell calculetion by means of the Cornpone
 

gk

progrem for the ORACLE which yielded sets of 3h-group disadvantage factors, e

one set for each region of the lattice cell, When employed in & 3h-group finite=-
\ eactor, Cornpone calculation, the correctly “disadvantaged" absorptions of each
element in each region of the reaetor Were calculated. o : | |

The disadvantage factor is defined by the following equation'

: f gav RS
e Te T

vhere _ :
e volume of lattice ceii‘

«E-»f-=' volume of region 3 in the cell

e S oS

= neutron flux in differential volume dV in
neutron group n.

3.9 Costs of Materials and Facilities and Interest Charges

The basic cost data employed in this study to calculate fnel cycle costs
are given in Table 3.1.- These data are believed to be representative of the
costs of MSBR materiels and amortization charges.

Fuel Stream.Processing The cepitsl cost of the fuel stream.processing
 plent was based upon & cost study by Welnrich,™l who estimated the capital
_charges for a plent to process continuocusly about 20 ft3/day of fuel salt. A
pdant of this cepacity is within the region of interest of this study. Weinrich's
data were reviewed by'Chemical‘Technology Diviesion personnel ‘for comparison with
, morejrecent-cost,data'and'coSt estimating practices at Oak Ridge National Labora=
tory and, as e result, his date were edjusted upvard. These date and the ORNL
revised figures are presented in Teble 3.,2. The ORNL estimate is‘apprOximately
tvice that of Weinrich. These estimates vere made from functional flowsheets
representing the best available design information on the fluoride volatility
and HF dissolution processes. ‘ '

' In optimizing MSBR systems to obtain the most economic combination of
fuel processing plant cost end fnel salt replacement cost, it vas necessary
to extrapolate the QRNL cost estimate in Table 3.2 to both smaller and larger

 

W. G. Stockdele, D. O. Campbell, and W. L. Certer.

g ;

X )

L]
al

i)

 

- 25

Table 3 l. Items end Basic Cost Data Included. ‘in the
Fuel Cyele Cost of & Molten Salt Breeder Reactor

 

Unit Velue Interest Rate

 

($/kg) _ (%/yr)_ _
Ureniten inventory ) | | 15,000 ' h
Thorium 1nven£ory (as ThFl;) | 27@') T .-12.7(b)
Fuel salt 1nvnetory (U excluded) | | h0.3(c) . 12.7(b)
Fertile salt imventory (Th excluded) ' | 1;,5.6("")_" | 12.7(P)
Thorium amortization (20-yr cycle) ; 27 - -

' Fuel salt replacement o  40.3

Fertile selt replacement (20-yr cycle) w6 2.6
Fuel strea;m che;nica.l processing pla.nt 29(6‘). .
Fertile stree.m chemicel processing plant SIS ELER D ati s 29(‘1)
Ereeding credi‘b 15,000 “

 

(2) Th value at $22/kg plus $5/kg for preparation of selt solution.

(b) Includes interest et 6%, income taxes at 4.64, and local taxes and
.:I.nsure.nce at 2.1% o L

(c) Based on LiF et $l&h/kg and BeF, et $15 ho/kg plus $11 kg selt for
prepe.ratien. Atomic concentra%ion of Li is 99 99% Li o

(d) Includes 14% 1nterest on capital investment plus 15% for opera.tion '
- maintena.nce. L | -

 
 

Teble 3.2.  Cost Estimate of. Facilities for Continuous
| - Processing of Mblten Salt Breeder Reactor Fual Stream. |
- f  Weinrich's Cost Estimate Compared with Revision Made by QRNL

' Weinrich's Estimate ($) ORNL Estimate ($)

 Tanks and Vessels . .
_(Core Salt Section) -~

 

 

Installed cost (dl 35 x cost)

Core salt hold tenks - 19,500 39,000
Core salt fluorinators 116,500 ' 33,000
UFg chemicel traps‘_; S 31,900 - 63,800
,UFk‘UFE reduction tover 8,600 17,200
Vibrators, filters, burners, etc. 5,000 10,000
HF dissolving tenk 24,000 18,000
HF eveporators ~1Ql;000‘ ,7180,500 -
HF condensing tower 23,000 46,200
HF sfiérage tank 41)300‘,' 82,6007
KOH scrub tower 3,500 - 7,000
Miseellaneous storage and utility tanks 20,000 - 20,000»
Sub-Total | o 29h 400 547,300
Instelled cost (=1.35 x cost) 397,440 138,800
Coolers (Core Salt Section) . .
UFB ges coolers ' ‘_‘33600 '3,600 B
Reduction tower vent cooler | 3,000 - 3,000 -
HF vapor desuperheater | | | : §;600- | '9,600‘
HF condensing tower vent cooler 6,000 6,000
' Circulating HF cooler. | 48,000 - 48,000
| Circuleting H0 chiller 2,400 2,400
Sub-Totel | 72,600 72,600
Insta.lled. cost (rsl .10 x cost) 79,860 79,860
. Vessels and Tenks
(Blanket Salt Section)
Blanket salt hold tenks 9,800 |
Blenket salt fluorinator '5;500 | Fertile streanm
UF chemiéal trap L 3,600 | | processing esti-
Sub-Totel 18,900

. mated separately.

Az b
T 22

»)

 

- 27 - -

Table 3 sl

. . VWeinrich's Estimate ($)

Coolers (Blenket Selt Section)

;UFB-cooIéf-«'

Instelled cost (=1.10 x cost) -
Miscellaneous Equiment

Punmps |

Agitetors

Filters -

Freon :gfrigeration system

Fuel reconstitution system

Electric heating furnaces

Pipe heating equipment -

Fé + Eé ges supply systems

Fé‘compressors '

Sub-~Toteal -~ |

Instelled cost'(ni.35 X cost) o

Sub-'l‘otal of installed cost of |
major eqnipment

Attendant'Facilities‘

Special 1nstrumentation .

,General instrumentation o

Panelboards and slarms
SubwTota.l . -
Installed cost (=1.40 X cost)

| Piping, Painting, Scaffolds

etc., Installed Cost
Speciel piping
General. piping( S
Equipment footings and foundations s(?)
Pipe insulation '

Continued

2,400 -

2,640

40,000

6,000

20,000
160,000

60, 000
148,000

- 60,000 -

20,000

20,000 |
534,000

720,900

1,226,300

76,

24,000

160,000

224,000 -

4,500

231,000
'138,000

8,000

dRNL Estimate (#)

65,700
6,000
30,000
160,000
80,000
148,000
- 60,000

20,000

20,000
239,000

B 796,-11»0_‘_ |

1,614,800

76,000 o
60,000
160,000
224,000

4,500
1,210,100
181,500
8,000

 
 

Equipment insulation . .

- 28 - |

Teble 3.2.

Weinrich's Estimate ($) :

- Continued -

 

Electrical distribution, lighting, etc. 144,000
Pa.in’hing( c) ; - 28,000
Remote operating equipment - .75, 000
‘Field testing and inspection 25,000
Operating and safety supplies 15,000
Fretgnt(d) 37,000
Sub-Total - - 725,000
TOTAL INSTALLED COST . 275,840
Co_ntingericy(e ) - 2'.17', 580
TOTAL DIRECT MATERIALS AND LABOR 2,393,400
Feee and Expenses
Contractor's field expense( 8) o 119,670
Contractor's overhead fee (b,e) - 359, 000
Engineering and design(i) 478,700
Purchasing and shop inspection_('j) 119,700

Fstimated Cost of
Additional Facilities

‘Sempling fé.cj_.']_.ities

Ventilation .

Weste rer&éval " '

Cells end buildinge 1,500,000
Leboratory o )
Mock-up cell

Crane

TOTAL ESTIMATED FLANT COST o 4,970,500

ORNL Estimate ($)

20,000
144,000

136,300

72,000

25,000

15,000

48,400

1,767,800

3,606,600

901,650

4,508,300

| 2,254,150

901',_660_ |
225,400 -

70,000

10,000

1,700,000

750,00‘0:

20,000

60,000

. 9:81‘93500

O

O

L¥I
-29 «

‘Teble 3.2. Contimued -
.- Footnotes

(®)gstimateq by Vetnrich as 254 of major equipment purchase price.
-Estimated 'by ORNL as 100% of major equipment purcha.se price.

(b )Estimated. es 15% of maJor equipmen'b purchase r 1‘33'

(c)Esti'ngated 'as'.3$ of ma_.,jozf equipment purchase price-.

(d)ns'timaiea . h% of major z-éqfiipu.nefit' -vpuz;chase' price--»

(e )Estimated. by Weinrich as 10% o:f.‘ tota.l insta.lled cost. B
Estimated by ORRL as 25% of toba.l mstalled cost. |

| (f )Estima‘ted by Weinrich as 5% of tota.l direc’c materie.ls and la.bor cost.

(8 )Sum of. contractor s field expense. and overhead fee teken by ORNL as 50%
of total direct mater:l.a.ls and 1a.bor cost°

(h)Estimated by Heinrich a5’ 15% of tota.l direct materials a.nd la”oor cost. )
(i)Estimated as 20% of total direct. materia.ls end labor cost.
(3 )Estimated as 5% of tota.l direct materials a.nd 1a.bor cost. |

(k)Weinrich ellowed $1,5oo,ooo for ad.ditional facilities ‘that might be
ghared’ with reactor operation. e S L

 
 

plants. - The extrepolation vas made by assuming that the capital cost is pro-
portionel to‘the 0.6 power of the processing_rate. This method of extrapclating

- cost data hes been found reasonebly eccurate when applied to the chemicel in-

- dustry as & whcle and to plants which process nuclear reactor materiels.

There is a limit, however, to the extrepolation in the region of low'proc-
essing rates beceuse at some low rete, vhich may not be well defined, it is
economic’to change from continuous to'batch'processing“methods;n‘In thie study
it was assumed that the lower 1imit of continuous ‘processing would oceur araund

ft3/day, vhich corresponds to e fuel cycle time of 75 deys. (The fuel stream
volume was constant at 530 £t3, ) When the fuel cycle time is T5 days, the ex-
trspolated cost curve (Fig. 3.1) indicates that the cepital investment is ebout
$5 million. ‘Furthermore, 1t was felt that the investment in 8, betch plant ‘would
not be sensitive to further increases in the cycle time; cansequently, the $5
million value wasflassnmed to epply to all plents_having cycle times greater than
T5 deys. A batch processing plant was estimated by Weinrich to cost $3.% million;
the ebove figure allows & pfemiumfiofi$l.6 million over Weinrich's estimate, o

| Fertile Stream Processing. The fertile stream is'precessed'only in &
fluoride volatility step and, therefore, requires much less equipment than the
~accompanying fuel stream processing plant. Wéinrich included the fertile
stream plent as en integral part of his fuel stream plant design and did not
neke & complete separate breakdown of the two costs. waéver, it was possible
to prepare a cost estimate for the fertile streem plant by extracting specific 7
items from Weinrich's.estimate end including_ellocations_for instnuments,_bnild-
ings; etc. ORNL pricing procedures were applied to prepare the.estinate given
in Teble 3.3. | |
This tebulation presents values that are appliceble to a plant processing
~ fertile stream at a rate of 20 ft3/day, the same basis upon vhich the fuel
stream processing plant was designed. For this rate it was estimated that the
capitel investment would be about $1.8 million dollers. These values were plotted
in Fig. 3.2; the remainder of the graph was obtained by'assuming the cost was
proportional to the 0.6 power of the processing rate. | . -
| CAPITAL COST (3x169 .

50

Y . ' . W 1 : . ) ’

UNCLASSIFIED
ORNL-LR-DWG 54794

 

5

T T 11T T T T T T T 71T

]
L L

L1111l

 

 

Confihu'ous Processing ——|—Batch Processing

 

 

o Lt | lll I L L Lttt 111
' 2

 

2 5 o - 5 1o o - 500

FUEL STREAM CYCLE TIME (days)

FIG. 3.1 CAPITAL cos1' OF FUEL STREAM PROCESSING PLANT FOR A MOLTEN SALT BREEDER
- REACTOR o

 
 

- 32 -

Table 3.3. Cost Estimate of Facilities for Contimuous
Processing of Molten Selt Breeder Reactor Fertile Stream

Tanks and. Vessels '

Selt hold tank
Fluorinator
UFB chemical traps
UFg’UFL reduction towers
Vibrators : '
UFB gas coolers
Reduction tower vent cocler
Puzps

Filters '
Agitstors
- Freon refrigeration

‘Bred matérial reconstitution
Enectric heating furnaces
Pipe heaters
F, supply
Fé compressor
Sub-Tbtal ;
Installed cost (-=1 35 x cost)

Atbendant Pacilities
Speciél instruments
" General instruments
'fhnelboards and alarms
Sub-Total |
Installed Cost (=1.40 x cost)

Installed Cost of Piping,
Insulation, Peinting, etc.

Special Piping

Genersl piping (élOO% of major equipment cost)

Ehnipment footings and foundations (=15% of
major equipment cost ) o

. Estimated Cost ($)

20, 000
20,000
43,000
12,000
10,000
3,600

3,000

2,900
6,000
23,500
8,800
20,000
8,800
2,900
2,900
225,400
304,290

40,000
30,000

© 15,000

85,000

119,000

3,000

225,400

o 33,800
"

o)

- 33 -
Table 3.3. Continued
Pipe- insuletion . -

Equipment insulation
Electricel distribution ..

~ Painting (=3% of major equipment cogt)

Remote operating equipment

. Fleld testing end inspection
* Operating.and safety supplies

Freight (<4% of major equipment cost)
Sdb-Tbtal

. Total instelled cost

Contingency (=25% of totel installed cost)
TOTAL DIRECT MATERIALS AND LABOR

’Contractor g field expense end overhead

(=50% of totel direct materials end labor)

Engineering end design (=20% of total direct
- materials and lebor)

Purchasing end shop 1nspection (n5% of total
-direct materiels and ldbor)

Additional Fecilities Shared
with Fuel Salt Processing Facilities

Sampling - e L En
Véntilat;on - | :

Celle end buildings

Ldboratofy

Mock=-up éellr

Estimated Cost ($)
1,200
3,000

21,000
6,800
11,000
3,700
2,200

—2,000
320,100

T43,400
185,800
929,200
46k, 600
185,800

46, 500

10,300
1,500
100,000
7,100
2,900

1, T«,200

 
 

 

: UNCLASSIFIED
‘ ORNL-LR- DWG 54793

g

O

Fl6.3. 2 CAPITAL COST OF MSBR FERTILE STREAM PROCESSING |
PLANT - |

 

ot rrertee e

10

w»
!

 

'CAPITAL COST(8x10")

ol " cosT-ozerexio (RATE)

 

 

 

G e e b
20 850 100 - 500

 FERTILE SALT PROCESSING RATE (ft”/day per station)
i

9

i

quantities by Eq 2. N

 

“ 35 -

4,0 REACTOR CALCULATIONS

h.l Reactor Size -
Fbr engineering reasone it was decided thet the MSER would be e cylinder

| having a height equel to the diameter._:

Core Size. In determining the core size of the ‘MSER 1t was necessary to
fix certain reactor pmoperties. -In this study'the thermal power, fuel stream
velocity in the core, end the temperature rise of the fuel in its passage through
the core were erbitrarily chosen._ The diameter cf the core ie related to these

 

e =[ e ]1/2\ @
s T
where“A, o | B
L el core thermal power per reactor
is_gstream velocity
| ~stream density

= heat capacity

.Fthf, filnfl |
n

ié_ fraction of core cross section occupied by fuel stream
: AT'w”’temperature rise ' S |

When the-apprqpriate~numbere -are substituted in this equation, & core
diameter of ‘ebout:T. 7 £t 1e obteined.f The power used in obtaining this dism-
eter was one=half of the total core pover for the. station, giving two reactors
for the instellation. This ‘agrees with the decieicn that the total etetion
load would nct be committed to a: single reactor. LT s | CLono

Hlenket Thickness, The cylindrical core of the MSER 1s surrounded by a
blenket on the sides end on each end. Based on previous studies, 13 the 4thickness
of the blanket was fixed at 3 ft on both ends &and on the side.. This.thickness was
sufficient to reduce neutron leakage to &’ tclereble level, As given in Teble 2.l,ifl

the side blanket is 90 vol - % fertile etream and 10 vol % graphite. o

Reflector Ehickness. The reflector vas chosen to be & l-ftuthick'block of
grephite surrounding ‘the eide end ‘end blenkets. The overaall reactcr dimensions,'
excluding the reflector, ere 13.66 ft diemeter by 13.66 ft high. |

 
 

4, 2 GNU Calculations |

The calculations for the MSER were made as indicated in the flow diagram.of .

Fig. 4.1, The basic mucleer calculetions were performed on the multiregion, one-

dimensional, 3h-group GNU programa for the IEM-TOk. The equilateral, cylindrical -

reactor vas treated as an equivalent sphere heving & ‘diemeter. 9% greater than the

L cylinder, the 34 groups of cross. sections consisted of 32 fast groups, &an epi=-

thermal group for the energy range 5. 5 kT-0. 6 ev,.and a thermsl group.

Inggt Date. Input data for GNU consisted of sPecifications of resctor
geometry, dimensions, and the homogenized atomic densities of the several ele-

~ments in the system. The concentration of each element was homogenized over the ,'

-region in vhich it eppeared. Since & small fraction of the fertile stream passes
through the core of the MSBR,'the;coreflconcentratiens-included the sum of fuel

- stream and that portion of the fertile stream concentrations;' The initisl vealues

of fuel, moderator, and fission product concentrations used in the GNU calcu-

- lations were based upon concentrations nreuiouslydeveloped'forfitheexperimental
gasacooled reactor (EGCR) and- from previous molten selt reactor'studies;i The
elements considered in these calculations are given in Table h l._

Output Data._ The GNU program provides & criticality search by which either
a dimension or one or more concentrations ere varied.until the multiplication
constant differs from unity by less then some small specified amount. In these
calculations the reactor was made critical by verylng the concentrations of

protactinium and thorium in the core. Since the thorium density in ‘the fertile~

stream carrier is Pixed by the salt composition, thie is equivalent to varying
the volume fraction of fertile stream in the core. Additional useful output

date were the fractions of neutrons involved in absorption and fission reactions

~ for each nuclear species in each region of the reactor.

4.3 Corngone Unit Cell . Calculation » - : - .
The second step in the muclear calculetions was to determine the atomic
concentrations of the heterogeneous reactor. (The GNU program could treat only
a homogeniaed system.) The coreIOf the MSER was visualized aS‘being composed
of & number of cylindrical, unit cells like the one diagrammed in Fig. k,2. The
unit cell contained six regions (see Fig. 2. 1): inner fuel zone, graphite tube,

annular fuel zone, graphite tube, fertile stream passage, and.graphite moderator._
- Wodetermine critical concentrations|—ae calculation to obtain j——ed to determine critical Thconcentra-

Wl

) . . ) . o i . . ' q -
‘ . . : _ . o IR . ‘ _ ‘I~—I

 

 

 

Préllmlnary GNU calculation _ccrn_panéfufni'-t cell | | | Cornpone finite reactor calculation

 
 
    

 

 

 

 

 

 

 

 

of homogemzed reactor. | |disadvantage factorsf | tionin core of critical heterogeneousi |
e e — = -1 reactor and cross section mtegmls

 

 

 

 

 

Output Dma o 1 Cost ophmnzanon to de’rermlne :

| Equ:hbrlum concentrations, { | minimum fuelcycle cost foreach§
> fraction, variable fertile stream — | inventories, neutron balonces—rl' parameter in ethbnum reactor S
|eycle Ilme and vanable ferhle -and processmg rafes S calculuhon oy
stream vqume. ST st . r Lo

-} EQuuhbnum reactor calculations| |
~ |{ERC-5 code)at variable poison |

 

 

 

 

 

 

 

 

 

 

 

Poison fracflon calculation . S Opflrmzed fue! cycle cost
|(PF-8and PF-9codes toobtain} | - |and fuel yteld e
|poison fractionas a functionof 4 , —
fuel stream cycle timeand
o fuel salt dnscard cycle. hme.)

 

 

 

 

 

 

 

 

 

Calculations repeated if |
output data indicate cross |
section integralsneed [—~"7"""""="
adjustment.

pme—- .-...—,...---_-_-.—_:---.--_-

 

 

 

FIGURE 4.1 COMPUTATIONAL PROCEDURE FOR MOLTEN SALT
BREEDER REACTOR |

UNCLASSIFIED
ORNL-LR-DWG 58644

"mm”mmwnm-—xufimm.fip“munumm-u,u-d.fim-—a—'- .

 
 

-38-

 

~ GRAPHITE MODERATOR
~ FERTILE STREAM ~—

 

OUTER FUEL TUBE (GRAPHITE)—

"ANNULAR FUEL ZONE ,
INNER FUEL TUBE (GRAPHITE)
 INNER FUEL ZONE—

re

s
rq
s

e

    
  

" INCHES =
| {.875 .
- 2.375
- 2.62

UNCLASSIFIED
 ORNL-LR-DWG. 58645

em
‘3.02

4.76
6.03

- '6.65
14.07

 Fig. 4.2 Unit Cell Configuration for Molten Salt -

.Breeder Reactor Core.

e

o
mi

- 39 -

Teble &.1. Elements Considered in Nuclear Calculations
- of the Mblten Selt Breeder Reactor

 

 Pa-233 - B Fissium
U-233 | . smas
v s

S U-235 | o ;_Graphite ;

- U-236 o - Fuel-Stream Carrier( e) )
o e 7 (63-37 mole § LiF-BeF,)
Np | - o FErtile-StreanrCarrier(b)

(67-18-15 mole % LiF-BeF,~ThF, )

 

() Uranium excluded in nucleer properties of salt

(b) Th content excluded in nuclear properties of salt

The cell vas examined using the Cornpone code9 for the ORACLE to develop & set of

34=-group disadvantage factors for each region of the unit cell. These disadvantage,
factors, defined in Section 3 8 were then used in subsequent Cornpone calculations
to determine the concentrations fbr the critical, heterogeneous reactor.

The Cornpone code treated the unit cell as an infinite cylinder heving zero
net current at the outer boundary Input information for the calculation was
the stream concentration in each region and the thickness of each region. The
streem concentrations used were those developed in the preliminary GNU calcu- _
lation. In eddition to the disadvantage factors, the code calculated the mnlti-
plication constant of the cell.‘ ;7_‘ |

k., h Cornpone Finite Reactor Calculation |

The disadwantage factors vere used in a finite reactor calculation using
the Cornpone program to determine the critical concentrations of the heterogeneous
reactor. As in the GNU calculation the reactor vas calculated in equivalent
spherical geometry.
 

-;ho'-

Input Data. - The. celculations were made on & 3-region model ~ core, blanket, ifi*’
and reflector. The concentration'of each nuclear species, which were obtained -

- from the GNU calculation, was homogenized over & region according to its volume

 

fraction in the region. The set (or sets) of disadvantage factors to be used.

with each concentration was specified as well as the dimensions of the region.

The machine calculation mnltiplies each homogenized concentration by the appro-
priaterdisadrantage facter so that all properties that are dependent on the atomic
density of that element are veighted by the relative flux to which the muclel are
exposed. The'concentrations are thns"“disadwantaged“ to reflect the heterogeneity
of the system. Fbr example, the absorptions in the i~-th element in the J-th stream
in the k-th region is computed‘by the ‘double summation

Aiyd: Z Z 1:33 (Au) Kis:lsk i J(Au) 3 (Au) Au AV o (3)
Vk o

where fik(au) is the group-mean qux in the homogenized incrementfiof'VOIume '
AVE, D is the disadvantage factor, N 1,3,k is the homogenized etomic con-
2

1,J,k
centration, and © (Au) is the absorption cross section.

It should be pointed out that disadvantege factors vere applied only to
element events (absorptions and fissions) occurring in the core. In the blenket
and reflector, element events were calculated as though the disadvantage factors'
were unity for all lethargy groups. '

Outpgt Data. The Cornpone program determines the fractions of absorptions
. and fissions of each atomic species in each region and the multiplication con-

- stant of the reactor. The code does not meke a "searck | on any of the input

7 information, hence it-is necessary to rerun the problem'with adjusted input if
~ the multiplication constent differs from unity by more than & prescribed small
 amount. In these calcnlations criticality was achieved by varying the thorium

concentration in the core.

Reaction Rate Coefficiente. FElement absorptions in each region of the
reactor were used to compute sets of reaction rates coefficients, i, J k’ which
are defined by Eq. 4. o ‘ |

| 1,J,k 4 |
1,5,k = N—ifi‘; = Z Z D, 3, k(lmt) (Au) g, (su) su av, . () |
n
L)

 

- 4] -

" The symbols have the same definition as given above for'Eq. 3. "Since the double

summation 1s computed by the Cornpone code, the calculation of Ci 3,k isrstraight-
yJ2 o
forward S S ' L '

| The coefficients are properly disadvantaged through the use of D ,J,
reflect the heterogeneity of the system. These integrals have the useful property
that, vhen mnltiplied by the stream atomic concentration and the volume frection
of the stream in the considered region, they give the fraction of neutrons in-
volved in ebsorption interactions with the i1-th element in the J-th stream in
the k-th region. Furthermore, if this fraction is:multiplied by the total number
ofrneutrons*born ‘per unit time;in.the\reactor, ‘the product is the absorption rate
by element 1 in stream J in region k. This latter quentity is very useful in
calculeting the equilibrium state: of the reactor a8 discussed below. :

Ine calculation similar to that described by Eq h, sets of fission rate
coefficients were developed for elemsnts thet hed a fiseion cross section. These

fission coefficlents were used in en entirely analogous manner to the absorption

coefficients to describe element fission events in the streams and regions of the

~ reactor. - A1l comments about the use of ‘the absorption coefficients apply to the

fiesion coefficient.

k.5 Equilibrium Reactor Calculations (mc-s)

Ehuilibrium Reactor Calculations vere next performed on the critical Cornpone

reactor by means of the ERC-5 coaer? for the IEM-T0%. This program integrated
the reactor with the fuel and fertile stream chemical processing systems and com-

puted pertinent equilibrium properties of the system.

Inggt Data. The equilibrium calculations required the following input in-
fbrmation°. fuel and fertile stream,volumes, volume fractions, process cycle
times, .process holdup times, end critical concentrations; reactor powver, poison
fraction override, fuel reserve time, end recovery efficiencies associated_with

.\fuelkand‘fertile stream,proceSsing, The ERC-5 codesol?ed'a-system of equations

based upon'conserration of mass, criticality,iand'conservation of neutronsj these
equations used the-absorption_and fission reaction rate'coefficients calculated
from the Cornpone data. All neutrons were accounted for, including those ebsorbed
in fertile materials, moderator, carriers, etc., and those leaking out of the

‘bleanket or lost as delayed neutrons.
 

- k2 -

-Output Data.. The program celculates theeequilibrium stream concentretions
and neutron absorpticns in both fuel and fertile streams for the elements listed

in Table 4.1. Also the inventories and mess processing rates are computed for

all uranium isotopes, thorium and protactinium. Since the sales philosophy is

to sell e product thet has the same composition as the system.mixture, the.
fractions of recovered fuel and fertile streams that are directed to seles are
calculated. Additional values calculated by ERC-5 code are the fraction of ,
fissions in the fertile stream and the inrentory of fissionable material reserved
for a possihle 30=day shutdown of the processing facilities._'

- The prcgram,ofTErs the option of atteining criticality'hy'addusting the U-233

- and U=-235 concentrations in the fuel stream or by edjusting the volunme fraction of
_thorium in the core. In these calculations the second option was employed. Slight

adjustments in the amount of . thorium in the core: hed & negligihle effect on the

carbon-to~uranium ratio and hence on the neutron spectrum

- In some instances it was desirable to specify the fraction of neutrons that
would be allowed for losses in xenon, fuel fission products, end leakage. This
condition could easily be treated on the ERC-5 code by specifying a fictitious
atomic concentration and absorption rate coefficient thet gave the desired ab-
sorptions. Xenon and fuel stream fission products were treated this way'because
their amounts are controlled by predetermined processing rates. Leskage 1s con-
trolled by the reflector design. " Lo e
a)

- 43 -

5.0 FUEL STREAM POISON FRACTION CALCULATIONS

5.1 Poison Fraction

The total fioison fruofionréénérated'bylfission produofs in & reactor in-
cludes the contribution to neutron losses from fuel stream Plus fertile streanm
fission products in both core and blanket regions. "Since ‘in the MSER the number

of fertile stream fissionsuis'afiomallfporfiion of the totel fissions end to sim-

plify the calculations, the totel poison fraction was aessigned to the fuel streanm.

- By definition,

e\ neutrons sbsorbed by i-th fission product in fuel stream
Poison fraction = ¥j[:. neutrons ebsorbed in U-233 *_3-235_;n fuel stream

 

i
g}: :13.(13."1"5.*’fla::.""i)‘r - sy
| oV o S
vhere, |

'»Ni;l = atomic concentration of i-th fission product, atoms/cm ,

.fii,l- .%. volume fraction of fuel stream.in core, fi' _ '. |

;1’2a.f= volume fraction of fuel stre&m 1n end blanket,

& "sl'average effective neutron flux in ‘core, neutrons/cm -sec,'
02;;;j:= average effective neutron flux 1n end blanket, neutrons/cm -sec,
7oifi -;fiurfieffective ubsorption cross ‘section for 1-th etom, cm?,

”t;. e % otal fission rate 1n reactor, fission/sec,,uu,”o*: -
_vfi'J "= neutrons born per fission, TR -

7 = meutrons bora per neutron sbsorbed 1n fuel,

V.. .= total fuel streamvolume,ém?.'_fjA e

The atomic ooncentration, Ni 1, can be expressed 1n terms of known quantities
by considering the steady state of the i-th 1sot0pe. Equating the production rate
to the sum of all removal rates, there obtains .

 
 

ek -

- % + N 1( 1, 1°1°i + fi 5 2°i) + o T (6)

The value of N, 1 from:Eq._écsnibe substitnted into Eq. 5 to obtain

 

1,
f?['h”'flfi lffi+&2217:f. ) °'_'1(5'
pf = v 'EI”'i' oo T
: o1 k + - fi 1¢ > +. fi 202 Yo Ti '

'Symbols not previously defined are

Y, = yleld of i-th isotope (for gome nuclides this number hed to
T . be edjusted to account for the existence of & precursor isotope
~"in the chemical processing schemes), |

A, = decay constant of i-th isotope, sec 1,, ' - ,
E, = efficiency of removal of i-th isotope in chemical processing, _

T, = cycle time for i-th isotope in chemical processing, seC.

The quantity E /T in Eq. T expresses the removel rate of the i-th isotope
in chemical processing In MSER processing, T assumes two vslues, jdentified
as T. and T a’ the values being characteristic of the chemical behavior of an

1 1
atom 1n processing. T, refers to those fission products whose removal is.

accomplished in the Hr}dissolution step (see Table 5.1); therefore, T, is the

| actual fuel stream.cycle time through the chemical processing plsnt.' Tid is
Vessocieted with those fission products wvhose removal is accomplished by dis-
carding e portion of the uranium-free fuel salt each time the fuel stream is
processed. The time Ti is indgggendent of the time T 13 there i, however, the

restriction that Tid must be greater than T1 In the economic csses, :ld will

be several times larger than ?l

The total poison fraction attributive to ‘the fuel stream is the solution
to Eq T. Through this equation the total poison frection is related to the
cycle timss Tl_and Tid and thereby to #he capital investment in the processing
plant end the replacement cost of the fuel salt. Furthermore, it is possible
:to optimize these costs for & given poison fraction by the appropriate choices

of T, and Tid. This optimization was made in this study.
)

 

- 45 -

5.2 Solution of Polson Fraction Equation

. The total fission product poison fraction was convenlently calculated using
PF-8 and PF-9 codes for the ORACLE which solved Eq. 7. Detailed knowledge of
cross sections as & function of energy ‘for the individual fission products was
not availleble; however, reasonsbly relisble thermal cross sections are known.

It was necessary therefore to relate fissionifiro&nct ebsorptions to ebsorptions |
in another element. for which more extensive eross section dats are evaeilable.
Carbon was chosen for the refErence element.

In Eq. 7 all of the terms sre known except the term ¢o."Erom-previous GNU '
or Cornpone.calculations e reaction rste_coefficient,_cc,,fbricsrbon cen be com~
puted as the quotient of total carbon ebsorptions &t all energles in a region
end the homogenized concentration of carbon stomsr(see“Seotion k.4). Using this
quantity an effective thermal flux cen be computed &8s |

 

S . | eff F% v C : ‘7141, B .

¢th = . . (8) :
i L s °b “Y/Pc,e " |
in wvhich oth'is'ihe‘thérmsl microscopic ebsorption oross’section for" carbon

end D 1is its thermal dissdvantage factor. The other qnsntities'vere~defined

above in Eq. 5. > _ -
» If it 1s desired to treat fission pro&ucts as l/v ebsorbers, it 1s only

' necessery to multiply both sides of Eq 8 by the thermal absorption cross

h eff th

: section, O : » to obtain the absorption rate. The oth 0, 50 obtained ‘may be
~ used in Eg. 7 in computing the poison frsction. On ‘the other hend, & more

pessimistic - but more realietic - computation is to include the resonance eb-
orptions end in some manner sdJust the thermsl energy cross sections to re-
flect these resonances. An efTEctiVG eross section vas calculated fOr each

Tfission product by including the resonsnce ebsorptions in the folloving manner;

e e o [v E ](RI’ o ;-«(9)
| ffiz o

 
 

 

= 46 -

| ,_-(RI_)i* =  resonance integrel for 1-th #u’élide, cma, | |
| Zf - mé.c‘.zl'gs;:o_lia_ic_. fiési_on éross .’seé't;.:!._c;h in :eac§§r, cm'l,
th :: slowing down power in-reacfior, c;._m;'l,
£ = fraction of totel fissions occurring at therma.l energy,

vV = mun'ber neutrons born per fission.

The terms v Ef: E Zt’ and f are computed by the GNU code for the IBM-?O’-L

Both sides of Eq 8 can be nmltiplied by 0' ff from Eq.. 9 to obta.in L
| F, v C, v 2: - ' e R
eff eff [ ] [ th ]
¢ c = c (RI) - {10)
th 1 : a§ v/p_ £t }3 |

- When the subscripts 1 and 2 ere inserted _to denote 'core region and, end blanket
region respectively,. .fwo expressions are obtained for insertion as the ¢o tex__-x_ns
of Eq. 7. These ere | - |

 

 

F_vC. v 2. CoT
-°§§f1“§fl - [cfh.,’r 1 fE' (R1)y ] e A
o ; e _V/Dc.- f.g___t-‘
vhich 15 substituted for the term °1 of, and
T grie s cn bl p s -
- ¢:-§f2 o-:ff. = _—-—--l—:h L2 [o-:h + —£ (RI)i] e {12)
2 LGBy, LY ey, M

vhich is substituted for the term 0,07 .

The 's‘oliltion of Eq. T revised 'by Eqsll a.nd 12 1s the desired -poison fraction.

v

The value of f

£e 2y

for this reactor vae 0.6013.

O
L
+)

1)

i 4T -

. Resonance Integrals. Veluee of resonance integrals have not been reported
for a1l hh nuclides of Teble 5.1, '
The ones reported by Nephew7 wvere used, end, for the unavailable velues,
assumed or calculated values were used. When & celculetlon vas made, the method

-for infinite dilution described byDresner17 was used.
5 3 Fission Prodncts Included in Poison Fraction Calculation

The fission products used in the poison fraction calculations vere those
recommended by Burch, Campbell, snd Weeren.3}strty-four nuclides that would make
an eppreciable contribution to the poisoning were chosen; these are listed in
Teble 5. l.‘ The isotopes of xenon are not included in this tabulation because the
poisoning from xenon (primarily Xe~135) is eo large that it is treated separately,
and & special processing method (gas sparging) mst be employed to bring this value{
within tolerable limits. HEnce the poison fraction celculated by Eg. 7 fbr the
fiseion prodncts in Table 5.1 excludes any'xenon contribution.

The hh fission products are divided into three groups which clessify the
elements more or less according to their chemical behavior in the systen. The
first group contains ‘the metels that are noble relative to nickel and might dbe
expected to be reduced and plate out ‘on the walls of the system. Also included
4in this’ gronp are the iodines and bromine that are probably removable by gas
‘sparging end hence may behawe like xenon. The noble metals and the halogens are
treated as if they are removed from the fuel solntion on & very fast cycle and
thus contribute little to the poison fraction.‘ '

The second group contains the rare earths that are removed hy precipitation
in the: HF-dissolution process and are thereby controlled by the fuel stream cycle
time. This is the time referred to above as fl : o

The third group contains the alkali and alkaline earth metals that ere soluble,
in the Hdeissolution process end are removed by discarding the fuel salt on a

specified cycle. This cycle time 15 identified above as ?ld
>k Gas Sparging and Effective yiela -

Fission product nnclides which are daughters of gaseous precursors will have
K effective yields that are smaller then their actual fission yield beceuse the
| gas sparging operation removes & portion of the parent atoms before decay. The
\F/ fraction of gaseous nuclides of & particular species which undergo decey before

being sparged is

T
 

~ Tgble:5.1.: Fission Product Nuclides Included in '

Poison Frection Calculations

 

Nuclide

' Thermel  Decay  Yield

| Resona.nc
Integra.lid) -

Atoms removed by p&ating on walls or by gas sparging

- Rh?163 i~ - :
0.064

Mo-95
Ru-101

Mo=97

 Ru-l02
~Ru=104
. Mo=100

I-131

- I-129

I-127
Br-81
Zr-93

zr-3)

. 'Aton;s

Gd-l57
Gd-Eu-155(°‘)
Sm-149 -

Sm-f-fli-lfl _ .

Eu-153
Nd-14%3
Sm=152
Pu-14T
Nd-145
Pr-141 =
Na-1k6
La=139

Nd-1
| Nd-l% La-mo(a)

Ce=lh2
Y-89
Ce=-140

- Nd-150

Cross 7+~ Constent

Section
(barns)

=822

-
U ovP

-

&P o

0.16x10°
0.7 x102
wosn@
7000
400
290
150
60

o By

OCOED® Om®

-

n:c:#*&flco4rebua

- (see

stable

Com

n

e
e

o

 o.995u0™
"iistfible;\-‘

1"
"

 0.666107

steble = .
0.1281x10°1
- stable- L
0.300x10°2
_stable
| o.mao'-'B
- gtable
o .0.28x10° ~T
0.106x10-T
. 0.67 x10-8

0.182kx10-2

1y

0.05

. 0.062
0.01+2
 0.065

0.029
0.0025

0 0013
0.065
.059

0.0033

0.0013

0.052
0.0021

0.015
0.029

0.056

0,06 -

. oms

o (b)

| 'o 063
2.0 OO5 A_f'”

(berns)

1000
101
131

- 12.2

26.7T

15.8
6.3

25

... 16T

- 83.2
- 43.9

removed by precipitation 1n HF-dissolution e |

. 0.000L
0.0003
10.007

e 64

0. 219x10

3315
1512

3T
2850

B o

C 310
- 16,

10(e)
ek, T

-?;_1o(c)a -

| tc)-
;0.184 )

1ofe)
)

.0

= 49 -

Teble 5.1. - (Contimued)

 

Nuclide - - Thermal . Decay . Yield Resonanc?
. ‘ Cross ' Constant ‘ Integra.l )
Section o 1 '
(barns ) (sec™) (varns)

 

Atoms removed by mel salt discard

Cd-113 0. 25::105 steble 0.0001 %1.9
sr-89 130 ~ 0.149x1076 0.027(b) 193
Ag109 - 8k - gtable  ° 0.0003 1396
rgl0r 3% " 0.002 198
Cs-135 15 . | 0. ooohs(b) 375
Se-82 2 " | 0.0025 0,347
Cs~137 2 0.732x109 - - 0.0308(®) . 37
Sr-90 1 stable 0.059 o 10le)
Ba~138 0.7 4 " 0.0Li4(b) 0.0021
Te~130 0.3 " 0.02 8.6

 

(e) Considered together because cross. sections and/or yields are sbout
the same. : _

(b) ‘Y:Lelds are edJusted to reflect" gas sparg:lng of gaseous precursors on

e 6-minute cycle.

(c) Assumed value of resonance mtegra.l since no da.ta. for ca.lculating
availa.ble.

(d) ] Except es indicated by footnote (e ), velues are from Nephew (reference 7)

‘or calculated by method. of Dresner (reference 17)

 
 

o 7\d..fsaoaa)r' _
: 2
Rdecay 7"s;parge '

vhere the terms designate ‘the decay rate. and the sparge rate. The effective -

yield then becomes |

Effective yleld = (actusl yleld) - ,de‘_’:yl - o (13)
o - : decay  sparge :
For example consider Sr-89, & daughter of Kr-89, under conditions for
which the averege sparging time of the fuel stream is six minutes.

kr-89 —W ) '99 "—1'5——9 R
e 0.693 - ”
Effective yleld of sr-89 = (0.048) — e = 0.0271

In thie exemple the effective yleld of Y-89 vauld be the same. R

Where applicable, effective ylelds based on & six-mipute sparge cycle vere
used in poison fraction calculations in thies study. : ' '
5.5 Fission Products as 1/v Absorbers - |

A series of calculations was made using the poison fraction code for the

GRAGLE to esteblish the poison fractions associated. with e large mumber of

combinations of fuel stresm cycle time, Ti, end fuel salt discard time, Tid'
The initial celculations were performed considering thé fission products to be

Afv ebsorbers, and the results are plotted in Fig. 5.1. The curves represent
“the solutions of Egs. T, 11, end 12 in which the resonance integral term, (RI)i,

has been omitted. Values along the abscissa of the curves have been divided by
eta so that the poison fraction is expressed as fission product absorptions per

neutron born.
5.6‘ Fission Product Resonance Absorptions Included in Poison Fraction Celculations

A second set of curves, Fig. 5.2, was constructed from the solutions of
Egs. T, 11, and 12 to neflect the influence of fission product reéonance abe-
sorptions on the poison frection. Resonance integrels of the individual fission
products given in Table 5.1 were used. At the same velues of T, and T, the

C:
)

 

effect of :anluding the resonance absorptions is to a.pprecie‘bly :lncrease the
poison. fract:lon over its value when thc fission :products were considered to be
1/v absorbers. ‘A ccunpa.rison of Figs. .l end 5.2 shows ‘that for compareble
cycle times the inclusion of resonance absorptions mcreases the poison fraction
by & factor of 2 5 3.

Ve.lues along the abscissa of F:Lg. 5.2 have also been divid.ed by etea :I.n
order to express the poison fracticn on & "per neutron born basis.

ST Use of Fige. 5.1 and 5. 2
Figures 5.1 end 5. 2 vere used :l.n optimizing the :f.'uel cycle cost a.t & chosen

| poison fiaction. Alcng & line c_f constant poison fraction in these f:lgu.res

several compatible valces of T, end Tl q vere chosen, and the total fuel cycle
cost was calcula.ted for each pair of values. The cycle time, Tl’ influences the
fuel cycle cost through the capita.l investment in the processing pla.nt; the fuel
salt discard cycle tinme, Tl a’ reflects the replacement cost of the fuel cearrier.
The calcule.ted fuel cycle costs were plotted as & functicn of the fuel salt dise
card cycle time, T‘.Ld.’ (Section 6.2.1) and the cpti:mm cost and corresponding

cycle t:lmes vere determined. ‘
 

L3

-

L

days

  
 

(sfop) IWIL 37040 WY3IHLS 1304

09 e Lt

05 06 0T 08 -

04
-s3-

S 27

26

    
 

1.8

L) -~

(s§0P) IWIL 30A0 WY IYLS 13N

e
 

 

N Oa.v..,s.s..f‘ ...osr.,e..s‘... T a _ - , v L
e, ”. . . : i gk i

_6.

60

. 58
30

56
28

54
26

52
2.4

| 5.0‘_- S
22

48
20

46
18

44
e

42
14

40
12

38
10

- 36

   

  

-00or © Bt A N NC

O o o
(sAop) IWIL 3TOAD WVIHLS 13N

  
 

34
66 58 60 62 64 68

84

82

. T -oe-b ® @B

_..mm._._

  
 

 
 

|B7G!.4 n. . ®N

o =)

~ (sAop) FWIL T10AD WVIYLS 13N

32 34 3.6 38 40 42 44 4.6 48

30

26

24
 

6.0 PuRAMETER-sTUDIEs'AND RESULTS

The equilibrium reactor wvas studied. to determine the effects of variations
in certa.in reactor. cha.recteristics on the nuclea.r ;perfo::mance and economics of
the eystem. The investi‘gated para.meters were*‘ ‘

1. _poison fraction in fuel strea.m
. 2, fertile stream cycle time
' 3.' fertile stream volume
b, value of resonence integra.l of Pa=233 |
5."’? velue of epithemal fission Cross section ef U=233
‘6. eddition of ZrF, to stebilize fuel salt

o The lest three items perhaps are not rightly cla.ssified as pa.rameters since they
‘are not independent cha.racterietics. However__ in the cases of the resonance in-

tegral end the epithermal Pission cross sectlon, the ranges of uncertainty in
mea.sured. values are sufficiently broad to have significa.nt effects on ree.ctor '

. 'performa.nce. Ttem 6 vas introduced beca.use recent fuel salt studies have in-
 dicated & need for Z.th to inhibit oxide precipitatien of fuel atoms.

The two major parametere in this stud,y were the fuel stream polison fraction

‘and the fertile stream cycle time. The remaining four items were examined for
-varia.tions in these two major para.meters. The studies were made on the equilib-

rium state of the reactor described in Section 2. 0 using the ERC=5 codelo for

the IEM-TOL. However, in making the calculations for several values of epithemal

g fission cross sections of U-233 (Item 5), it was necessary to esteblish new
o critica.l cond.itions for the reactor using the GNU coc'!.e8 before the I'IRC-5 calcu-

lations could be ma.d.e.

: Ree.ctor properties thet were held constant during the parametric study
were the i‘uel stream volume (530.2 ft3) end station power (2364 th) ‘The

_effective carbon-—to-ure.nium ratio was calculeted for each equilibrium rea.ctor

e.nd varied only slightly from case to case because of slightly different equi~
librium conditions . The renge of C:U ra.tios for ell of the ca.lcula.tions was '

- 5020 to 5230‘ These values are the ectual C:U ratio divided by the therma.l

energy disedventege factor (0.879) for ca.rbon in the core. The verie.tion from -

case to case vas caused by small changes in the volume fra.ction of fertile strea.m

in the core for cha.nges in f‘uel stream poison fraction and, fertile stree.m cycle

 time.
 

In the equilibrium calculations:it waseassumed that the absorption and
fission rate coefficients.were not epprecisbly affected by small changes in con-
centrations (~ 10%) in importent elemente such as U-233, U~235 and fissium, and
by mach lerger changes (~ 1000%) in minor elements such as U-234, Pa-233, Xe,
etc.. In cases in which the equilibrium celculations significantly changed the
concentration to the extent that the reaction rate coefficients might no 1onger
epply, it was necessery to repeat the Cornpone unit cell and finite reactor calcu=-
lations with new'concentrations to develop new sets of reaction rate coefficients
(see: Fige k1), |

Several items in the neutron balence were specified for ell of the perametric
studies. These were the meutron losses to corrosion pro&ucts,“delayed neutrons,
leskege and fuel processing. The values'adcpted for theSe“Quantities vere re-

- spectively 0. 0008, 0.0043, 0. 0016, end 0 0022 neutrons lost per neutron absorbed
in fuel. S e -

- Corrosion product losses vere estimatedhrrdm the'equilibriumflccncentration
of corrosion products of INOR-8. _Delayed neutrons were calculated by the method
of Walker.%g Leakage losses were estimated from design considerations; it was
felt that this number would be small because of the emall emount of fissioning
in the blenket. Fuel processing.losses were discusseduabove‘inSection_3.6,

6.1 . Results of Eguilibrium Reactor Celculetions

PErtinent characteristics of the MSBR on which equilibrium calculations
vere: made are listed in Teble 6.1. - : '

Representative results of the equilibrium reactor calculations ere given
in Tebles 6.2 and 6.3 in the Appendix. These results include the equilibrium |
atomic’ concentrations of- major isotcpes in the fuel’ end fertile streams, & = =
neutron.balance fbr the system, the contribution of indivi&ual items to the
fuel cycle cost, the ‘volume - fraction of fErtile stream 1n the .core fbr the
Just critical reactor, and seversl) items of lese significance such &5 the frac-
tions of each stream sold as product end the fraction of total fiesions occurring f:'
in the fertile stream. T " |
 

 

+
4

Teble 6.1. Characteristics of MSER -

‘Ehermalpower, Mot - e e 23611-
No. reactors in stetion =~ -+ = - e S g
| Thermodynamic efficiency = L T R 0.k23
- Fuel stream volume per reactor, ft3 S e 2_65.1 S E
‘ Fertile stream volume per reactor, :E‘t3 S 30000
- Volume fraction fuel stresm inecore 06 1
~ Volume fraction fuel stream in end blanket 0.16
- - Volume fraction fertile stream in silde bla.nl:et o 0.90
o Volume fraction fertile stream in end blanket o Loh
. Volume fraction grephite in side blanket . 00 0
Volume fraction graphite in end blan.ket .00
‘Fuel stream holdup time in reprocessing, days | 1
TFertile stresm holdup time in reprocessing, days 1

The option used to achieve criticality in the ERC-5 ca.lcula.tions vas the )
va.riation of the volume fraction of fertile stream in the core. Since the
volume fraction of fuel stream was fixed at 0.16, ‘the sddition or subtraction
of fertile stream was made at the expense of removing or adding moderator. R
Cmsequently the Cc:Uu ra.tio in the core va.ried sl:lghtly :E‘rom case to case. How;- ,.
ever these sma.ll va.ria.tions in C: U ratio did not significantly affect the neutron
spectrum

6.1.1 Fuel Cycle 'I‘imes -

- The fuel stream cycle times reported :I.n Tables 6 2 and. 6 3 ere the. optimized
cycle times. . Eh_oh time has been so chosen that it. refleots the most economic =
- rate for the chemical proce'osiog for the chosen values of poison fraction and
fertile stream cyele tize. The fuel salt discard cycle time has also been
optimized. o ' ‘

In these two tebles the results ere for reactor systems in which the fission
product resonénce absorptions were included in the poison fraction celculation.
‘Results of calculations in vwhich fission producte vere assumed to be 1/v absorbers
ere not included because these are just optimistic special cases of the resonance
- 59 -

alisorption calculetions. For the resonance ebsorption cases the fuel stream
cycle time end fuel salt discard times are respe_ctijvely in the ranges.12.84 dsys
end 145-1550 deys for -poison fractions from 0.02 - 0.065 neutrons ebsorbed in
fission products per neutron ebsorbed in fuel. For the l/v sbsorption cases,.
the corresponding cycle times are in the ranges 12. 5-735 days and 400-8100 days
for poison fractions from 0.011 - 0.065.

| The procedure for determining the oPtisnm fuel cycle times was referred to
ebove in Section 5. 7 s.nd is discussed further in Section 6. 2.

~ 6.1.2 Neutron Balence -

Resonance Ahsorption Cases. ‘A portion of each of Tebles 6.2 and 6.3 shows
the distribution of neutron ebsorptions in the reactor. Examining the neutron
balence of Table 6.2 for increasing poison fraction, one finds that losses to
~ protactinium fall sbout 109 partly because the Pa-233 ‘concentration decreases

sbout 7% due to decreasing breeding retio. The decrease in losses to Pa is also
partly due to & decrease in the volume fraction of fertile stresm in the core by
‘ebout T%. This is significant because ebout 60% of the captures in Fa-233 occurs
in the cere.’ A similer effect is observed for the tabulations of Table 6. 3 for
~ other fertile stream cycle times.

Concurrently neutron losses to samarium and other fission products incresse
by ebcut 0.045 peutrons. This is more than half of the breeding gein and results
in more than one-half the rroduction of excess “fuel. . The longer cycle times at
the higher poison fractions effec‘t less purging cf higher uranium iscrtopes. The
| consequent ‘build-up of U-235 causes & decrease in the mean n of the system by
gbout 0.004. Neutron losses to U~-236 and Kp-237 increese ebout 1. 5-fold end

O-fold, respectively, or by abaut 0.006 end 0. 002 neutrons. SR -

l/v Absorption Ceses._ 'i‘he neutrcn bala.nces for these calculations are not

. | presented in this memorandum 'beceuse they are of limited interest. Since the

- range oi‘ poison frections (0. 011 - 0.065) is greater the.n that covered in -the -
resonance a.'bsorption cases, more varietions might be expected in the elementel
ebsorptions o The explenation of the trends s hcwever ’ :I.s the ‘same as given ebove.
For example 5. Pa-233 ebsorptions decrease gbout 14¥ over the ‘range of poison -
fractions because its concentration decreases sbout T4, and the volume fraction
of fertile etream in the core decreases sbout 13%.
 

Losses to samarium and other fission Pproducts increases by O. Ofih neutrons -
consuming sbout T0% of the breeding gein. Higher isotopes of uranium build up
because of slower processing rates at therhigher poison fractions with the.ao- -
companying'decreaSe-(~V0 005) ‘in the mean value of~n; Neutron absorptions by
U-236 and Np-237 increase by ebout 0.00k end 0.C 005 L

| 6.1 3 System Inventory

The inventory of fissionable materials, which includes U-233, Uh235, and
Pa-233, fOr the equilibrium.reactors is presented in Fig 6.1; & detailed break-
down of the inventory is given in Tablee 6.2 - 6.3. For both resonance absorption
and l/v absonption cases the inventory does not change very fast with poison
| fraction.over most of the range of poison fraetions.,.fibwever, atuloW‘values of
the poison fraction & sharp upturn in the inventory is cbserved. This occurs

because of the increased holdup in the chemical processing plant at these fast o

'processing rates. The proccessing rate. for the resonance ebsorption cases at
& poison fraction of 0.02 is about -equal to the rate for the 1/+ absorption cases.
at e poison fraction of 0.011.

The largest effEct-onffissionabIe inventOry?is observed for the variation in -

fertile stream cycle time, As the fertile stream cycle time increases from 35-200
days, the totel fissionable inventory inereasesIEbout 50%, or from around 860 to

_ 12&0 kg. . The increase is. ettributed elmost entirely to. increase of U-233 inventoryL

in the fertile stream.which rises over 5-fold.. Uranium—233 inventory in the fuel
stream decreases about 6%; concurrently the U-235 inventory increases ‘about 7%

'Inereased fissioning in the blanket at the longer cycle times causes the critical -
mass of 09233"in the core to decrease. However since the breeding gain also dee._:

creases for increasing fertile stream cycle time at constant poison fraCtion,
the purge rate of U-235 becomes smeller because less U-235 is routed to sales,
hence the U-235 ‘inventory in the core builds up. o ce R

- The thorium inventory for this series of calculations was maintained constant
at 270 tomnes. | |

'fProtactinium~233'inventory is not very sensitive to changes in poison'fraction'

.or fertile stream cycle time, Since Pa-233 is not-refioved from the system in the
| fluoride'tolatility process, it builds up until“its decay“rate is exactiy”eral to
the U-233:production rate, Therefore. the Pa-233 4inventory will change in direct
.proportion to the breeding ratio.. For this system the inventory is in the range
100-110 kg, o | '

(‘\
e UNCLASSIFIED
CogmElee ~ ORNL-LR~DWG 88647

     

 

“"POISON FRACTION
 

-62 -

»6.1 h Fuel Cycle Cost

A breakdown of the fuel cycle costs for representative cases is given in
Tebles 6.2 and 6,3 These costs were calculated using the basic cost data given
.in Table 3.1. o L ' ' o ' |

o The largest single contribution to the fuel cycle cost is the charge for o
.the fuel processing plant which contributes up to 40% of the total cost. At
.the faster blanket processing rates, the fertile stream processing plant cost
also becomes & significant part of the total cost, gt & 35-day blanket cycle
ftime about 30% of the cost may derive from fertile stream processing '

_ Total inventory charges on fissionable materials, thorium, fuel carrier and
'thorium carrier account for 40 ~ 55% of the fuel cycle cost. Individually, the
-thorium carrier (~ 200 tonnes) contributes most to the inventory charges, from |
-,15 - 20% of the. fuel cycle cost; fissionable inventory contributes about 7 - 13
néthorium inventory contributes 12 - 16¢; and fuel carrier inventory contributes
'wemwaomyfim15-25n | | -

| Thorium amortization and thorium carrier replacement charges, which are 3
jamortized at 2, 6%, are not an appreciable portion of the fuel cycle cost, being
only 2 - h% of the total.' On the other hand, fuel carrier replacement charges

- ,become a significant factor especially at the lower values of poison fraction

 

;because of the high salt discard rate. At very high poison fractions (0 065),
rthis contribution is only about h% of the total fuel cycle cost, whereas, at low
‘poison fractions es much as 25% of. the cost is due to salt discard. 7

IR Breeding credit is an item of the fuel cycle cost that is directly prOportional
’to the breeding gain and fissionable inventory. The high erl yield reactors '
(6. 8¢/year) have breeding credits of about 0. 13 mills/kwhr, in the very highly
| p01soned reactors (fuel yield G?l%/year), the breeding credit is only about
0.0k mills/kwhr.‘ Although allowing the fertile stream cycle time to increase
- from 35 to 200 days lowers the breeding credit through increasing the fissionable
,,inventory, the effEct is not so pronounced as that caused by allowing the poison
;,fraction to increase. 7' . | | |
f:6 2 Poison Fraction Studies in Which Fission Product Resonance Absorptions are
| Included :

o A series of equilibrium reactor calculations was made for a range of fuel |

‘fstream poison fractions from O 02 - 0. 065 neutrons absorbed in fission products
- 63 -

 per neutroniabsorbed in'fuel.. This range of poison-fractions was epplied to

| fertile stream.cycle time parameters of 35, 50, T5, 100, 150 and 200 deys. The

L range of poison fractions was esteblished after a feW'preliminary calculations
to: include reactors with quite favorable breeding potential as well as those
| that are approximately “hold-your-own systems. Poison fractions lower then

- 0.02 were not considered because the required fast processing retes result in

‘_high fuel cycle costs without appreciable incresse in breeding gain. At & poison -
'fraction of 0. 065, the MSBR shovs & small, positive breeding gain, at higher - -
poison fractions it is doubtful if the system will breed.

These two parameters could very conveniently'be treated in the ERC-5 code
sinee they are items of input deta. The poleon fraction ae such does not appear B
,in EBC-5 input; however, the desired poisoning effect can be obtained by using

. fictitious fission product concentrations end fictitious reaction rate coefficients.
The net effect of edditionel poisons is to decrease the breeding gein and corr-

'sponding breeding credit. For each combination of poison fraction and fertile -
:; stream cycle time the code calculated equilibrium atomic concentrations, in-
:Q’ventories, neutron absorptions by elements, thorium concentration in the core

n'and processing rates., .
- 6. 2.1 Fuel Cycle Cost thimization ;

The fuel cycle costs were optimized for each value of the fuel stream |

o poison frection and each value of the fertile stream cycle time.r For each com-
':-bination of these two parameters the fuel stream processing cycle which gave :
the lowest total erl cycle cost vas determined., The procedure is discussed
'helow. | o o | L o

| Fuel Yield Versus Pbison Fraction. Fuel yields calculated for the equilib- )
r,rium reactor were plotted as 8 function of the poison fraction (Fig.,6 2).

'rtnearly 1dnear relationship indicates that the fuel yield is inversely propor-‘_‘

tional to the poison frection. The plots begin to curve in the region of low '
,e:ppoison fraction because the required fast chemical processing rates begin sig-,hii:
;,fnificantly to increase the fissionable inventory through holdup in fuel proc-
| jessing The result is & lowering of the fuel yield. The effect of increasing
fertile stream cycle time is to decrease the fuel yleld for a given poison
-fraction. This occurs becanse of increased fissioning in the fertlle stream and
'the accompanying increase in fission product conoentration plus an inrentory

increase.
 

 

UNCLASSIFIED

=64 - |  ORNL-LR-DWG 56649

 
B“

Fuel Salt Discard Time s a Function of Fuel Cycle Cost. As discussed
in Section 5.0, each value of the fuel streanm poison fraction can be atteined
by operating the reactor at several values of fuel stream cycle time and fuel
sélt discard cycle time, and there is some combination of these times for which
the fuel cycle cost is e minimum. - For each selected value of the - fuel stream ,

'poison fraction, several pairs of compatible values of these two cycle times

were chosen from Fig. 5.2, and the total fuel cycle cost was celculeted for each -
peir of values. The fuel cycle time determined the capital investment in the

-processing plant; the fuel salt discard cycle time determined the replacement

charges for the fuel salt. A plot of fuel salt discard cycle time versus fuel
cycle cost at constant poison fraction geve the curves exhibited in Fig. 6. 3.

The minimum of each curve represents the point of most economic operation; the
corresponding fuel cycle cost end fuel galt. discard cycle time are read directly.
¥When the optimum fuel selt discard cycle time is entered into Fig. 5.2, the
optimum fuel stream cycle time is fbund.j" | '

thimnm velues of the fuel . cycle cost, fuel stream.cycle time and fuel salt
discard cycle time ere given in Tables 6 2 and 6 3. -

6 2.2 Economic PErformance

| The optimized fuel cycle costs obtained'by the ebove procedure have ‘been
plotted as =a function of the fuel yield in Fig. 6. L, The curves have been calcu~
lated for & plent operating 80% of the ‘time, The most fevoreble fuel ylelds are
obtained at the shortest fertile stream cycle time; however, the corresponding
fuel cycle costs are high. LikeWise'the fuel stream.must be processed at a
relatively fast rate as indicated.by the lower values of the poison fraction.

. Fuel ylelds of the order of T%/yr can. be atts,ined at a fuel cycle cost of eround
S 1. 7 mills/kwhr. - e R e |

In the lower range of fuel yields it appears that fuel cycle costs of O. 75

%o 0.80 mills/kuhr cen be attained at fuel yields of l to 2%/&r. ~These conditions

require fErtile stream cycle times of 150.- 200 days. The curves could have been
extended in the lower regions of . fuel yields by perfbrming calculations at higher

',velues of the poison:fraction, however, uncertainties in basic data, .g., cross_

sections and resonence integrals, would lend doubt as to whether such & system

would heve & positive breeding gain.
 

 

UNCLASSIFIED
ORNL~-LR~-DWG 358850

 

Ji

+
i

b

YRR

B 4 e b =
Ve ir e
e e e esl e e

e aes bl e

 
 

——
W

 

 UNGLASSIFIED
ORNL -LR - DWG 5865]

 

 
 

| | In this series of calculations fuel stream cycle times ranged from 12 ~ 8h
days, and fuel salt discard cycle times ranged from 1k5 - 1550 days for the case

_ vthat fall on the envelope of the curves

The dashed envelope curve has been drawn to indicate the estimated maximum
,performance of this reactor. It might be possible to extend the envelope of the
"family of curves out this far by modifications in the C:U ratio in the core and
by optimizing the fertile stream cycle time. These refinements to ‘the calculations ,
'ljwere not made in this study, nevertheless, it is believed that the . chosen C U ratio
is near the optimum o

_ Along a line of constant fertile stream cycle time, the fuel cycle ost drops
Qirather sharply from its maximum value principally because of decreased charges on

- the fuel stream processing pdant at the longer fuel stream.cycle times and lover.
o ffuel salt repdacement charges for ‘the longer salt discard cycle time. During this p

 

‘ linitial drop in fuel cycle cost, the breeding credit is also decreasing, but the -
f”initial loss of breeding credit is far overridden by the savings on the processing :
‘_'plant end fuel salt replacement mentioned above. Consequently the initial drop in .
fuel yield is not as fast'as fbr'the fuel cycle cost. Eventually, though, as the

3 processing time becomes long (increasing poison fraction) the savings on the proc-w
:__essing plant end fuel selt discard are not so effective in lovering the fuel cycle :

ri:’cost and the rate of decline decreases. Meanwhile decreasing. breeding gain ac-
"1celerates the loss in fuel yield. Ultimately et higher poison fractions than _

- shown. on the graph & complete loss of. breeding gain would necessitate rising

"ffuel cycle cost because fuel would have to be purchesed.

7 Along 8 line of constant poison fraction in Fig. 6.4, the fuel cycle cost
'is affected principally hy'changes in capital charges on the fErtile stream proc-
| essing plant and ‘in fissionable inventory in the fertile stream “The contribution
7 of the fertile stream processing pdant to. the fuel cycle cost decreases with in- -
| 'creasing cycle time vhile the inventory charges increase. Initially the savings
"”on the processing plant overweigh the increased inventory charges resulting in

‘ “a net lowering of the total fnel cycle cost.‘ At the long cycle times, however,

o the inventory charges overbalance the lower plent costs and ‘the fuel cycle cost

a-reaches a minimum and begins to rise

| The decrease in fuel yield along 8 line of constant poison fraction is not
large and is due primarily to increasing inventory of fissionable material in the
_'system There is also the adverse effect that higher U-235 concentrations at the :
 

6.

flonger fertile stream cycle times have on the mean.value of 1 - an effect that |
lowers the fuel yield through lower breeding gain.

6 3 Poison Fraction Studies in which Fission Products were Considered to be
-1/ Absorbers , o
A series of equilibrium.reactor calculations was’ msde for & range of fuel
; stream poison fractions frcm 0. 011 - 0. 065 neutrons absorbed in fission products
- per neutron ebsorbed in fuel. This range of poison fractions was epplied to
fertile stream cycle time parameters of 35, 50, 75, 100, 150, snd 200 days and :
‘is broader than that considered in Section 6.2 for the resonance absorption cases.
When'fission products are treated as 1/v absorbers, the poisoning effect is not B
‘&5 grest as when the resonance ebsorptions are included, and it is therefore
possible to extend the range of calculations to lower poison fractions before
intolerably short fuel stream cycle times are reached. -

The equilibrium.calculations vere made‘using the ERC-5 code for the IBMuTOh
by varying the fertile stream cycle time and by using fictitious fissium concen=-
~trations ‘end reaction rate coefficients es mentioned in Section 6.2. These calcu-_;
-lstions were performed et the beginning of this thorium'breeder study before a
f_compilation of resonance integral dsta ‘became available, and it appeared that
treating the fission products as 1/v absorbers was the best approech to the
,problem,' Therefore, the results reported below should be considered es an
optimistic upper limit to the fuel yield and &an optimistic lower limit to the
-fuel cycle cost. | | |
| 6 3.1 Economic PErfbrmance ’: il_'f[fflri_i” L ,d:','_r(
| The optimized fuel cycle costs obtained by the above procedure are plotted
" es a fhnction of the fuel yield in Fig..6 5 These fuel cycle costs and fuel
'yields should be regarded as ‘rather optimistic values since considerable neutron t:.
ieconomy resulted from the assumption that the fission products behaved as l/v
?absorbers. Consequently it 18 believed that these curves represent a lower
bound to the MSBR fuel cycle costs; more realistic performance is that repre-
sented by Fig 6 h in which the best evailable resonance - absorption data were o
included. |

Fbr the conditions of Fig. 6. 5, fuel yields as high &g about 8%/yr at &
‘fuel cycle cost of ebout 1 3 mills/knhr vere obtained; a minimum fuel . cycle
cost of about 0. 68 mills/kwhr vas obtained et sbout h%/yr fuel yleld. The

 
 

| 0 -

UNCLASSIFIED

ORNL - LR-DWG 58@52

 

o
 

curves rise lineerly after attaining-their minima because of the influence of

the constant charge for the batch‘fuel;streamfprocessing plant. Batch operation
becomes effective in the region of fuel yields of 5 = 6%/yr at poison fractions o
of0025-003. | SR : |

The behavior of ‘the curves for variations in poison fraction and fertile
stream.cycle time can be explained by the same comments ‘made above in Section
6.2.2 and will not be repeated here.;_ ‘ o - o -

'6 L Effect on Reactor Performance of Varying Thorium Inventory _'

In order to study the breeding performance of the MSBR over a wide range of
operating conditions, the thorium inventory‘was varied in the range 100 - 400 tonnes -

- in & few representative calculations fbr which the fertile stream cycle time ‘was

35 .days, the fuel stream.cycle time was 20 days, and.the fuel selt discard time
was 1000 days. The thorium inventory was varied hy adJusting the fertile stream

| ~_volume in the range 2000 9000 ft3 per: station.L This particular series of calcu-

_:lations was. made at an early stage of the study, and the comhination of cycle times

is not optimum.with respect to fuel cycle costs at the various fuel yields.i How=-

'ever, ‘the dependency-of fuel yield and cost on’ thorium inventory is only weakly

v

affected, 1f at all, by choice of cycle times. Therefbre the behavior exhibited
by the selected cases is typical and may'be used as a guide in selecting thorium
inventories. The important results of these calculations are given in Teble 6.k,

Table 6 h Dependency of Fuel Yield and Fuel Cycle Cost on -'
B Thorium Inventory in a Molten Salt Emeeder Reactor “_'

 

 

Thorium Inventory .- Relative Fuel Yield . . Relative Fuel Cycle Cost .
(tonnes) -7-, oo T e T o
00 e 100
oo a2 18
7180 o et g e W e -

 

 
 

-

As.the;thorium-inventoryxincreases, losses,to;Pa-233»decrease, and there.
are gains in respeCt to mean eta and;U-236 abSorptions.' Breedingsgain increases,
‘but at a decreasing rate. lMeanwhilelfuelfinventory-in”the fertile stream rises.
As a'result,'fuel yield rises rapidly at first, and‘then more slowly as the in-
'fluence of increasing inventory overrides that of breeding gain. The cost'in-
creases steadily, however, being driven upward by increased charges for thorium
and uranium. The fuel yield reaches & point of negligible improvement at 270 i
tonnes of thorium, and this thorium inventory was used for further studies re-
ported above in Sections 6.1 = 6.3. 3 | R . o

, One hundred tonnes of thorium is not sufficient to £ill the blanket of the
‘reactor used in this study when the blanket thickness is 3 ft. In the corre-
sponding calculations, no adJustment was made for the greater leakage ‘that would
~result from a thinner blanket. Thus for the cese in the above tabulation, the
‘fuel yield should be less and the fuel cycle cost higher than calculated. " Ac-
cordingly the lho-tonne case was selected as a representative low-thorium case
for further study. ‘ o

Fbr the lho—tonne thorium fertile stream, a series of calculations was made
.to 0ptimize the fuel cycle cost and the fuel yield at a representative fertile -
stream cycle time (50 days) A cdmparison between these results and those of |
the- corresponding 270-tonne thorium case are presented in Table 6 5

The results of Teble 6. 5 were obtained for optimized fuel stream cycle
times and fuel salt discard times, and in 811 calculations the resonance ab- .
sorptions of fission products were included in the poison fraction calculations
The two cycle times are longest for the low fuel ylelds and shortest for the
high yleld cases,__Although_some_slight_trend with fuel stream:cyclertime is _
- observed, -the rule can be formulated that doubling the ‘thorium inventory adds = -
about 1.9 /yr to the fuel yield and about O. 2 mills/kwhr to’ the fuel cycle cost
regardless of the fuel stream cycle times.

The performance of the MSBR containing 1&0 tonnes and 270 tonnes of thorium
is graphed in Fig. 6.6. The solid curve vhich is drawn through the calculated
points is the envelope curve of Fig. 6.4. The dashed curve is an estimated curve,
based on the few 1L40-tonne thorium cases, for. the maximum.performance of the MSBR
at this low thorium inventory. The solid outer curve was then dravn to indicate the
 

,_.73 )

Table 6.5. mlten Balt -Bre'eder Reactor -

Dependency of Fuel Yield and Fuel Cycle Cost on Thorium Inventory
with 50-Day Fertile Stream Processing e.nd Optimized Fuel Stream
Processing Cycle Times _ |

 

Thorium Inventory, tonnes

270 ik | | 270 140

 

 

 

 

Fiel Yield, ¢/yr .  Diff. .  Fuel Cycle Cost, mills/kvhr . Diff,
20 03 1T . 085 0.63 0.22
2i5 0.7 1.8 ' 0.8k 0.63 0.21
3i6 = 1.7 1.9 - . 0.87  0.66 . 0.21
46 27 19 095 - 015 0.20
56 2.7 1.9 0.9 0.75  0.20

6.6 N B R B 1.37 ~0.20

 

estimated 1limit of maximum performance of the MSER when the thorium inventory is
optimized with respect to fuel yleld. The outer 'cui've_' also assumes thet a slight
improvement in the reactor can be found by & slight varistion in the C:U retio.
The C3U ratio (~ 5000) 4in this reactor was not optimized, but th:h velue is be-
lieved to be near the optimm '

The lho-tonne ‘thorium case is alee plotted in Figs. 6. T and 6.8 for optimized
fuel cycle times and for & range of ‘poison fractions from 0. 011 - 0. 065

6: 5 Effect of Velue of n-2§3 on MSER Performance

Uncerta.inty in the measured. values of the epithermal fiesion Cross sections
- of U-233 cen ca.use considerable variation in the celculeted performance e:E‘ the '
| MSBR, depending on the eet of Cross section va.lues that is used. This is the
case beca.use a.pproximately 30% of total fissions occurs et epithermal energies .
Réported epithermal va.lues of §-233 apparently a.gree within a.'bout t 10% of an
‘average or recommended“ set of values.h

 
 

 

FUEL CYCLE COST (mills/kwh)

0.9}—
0.8

Flg 6.6 Performance of a Molten Salt Breeder Reecior

T4 - UNCLASSIFIED

 

STATION POWER:1000 Mwe

PLANT FACTOR: 0.8

ORNL~-LR-DWG 59090

 

 

 

 

 
  

" 'ENVELOPE OF CURVES [
FOR THORIUM INVENTORY

 

OF 270 TONNES ==~ , /

    

 

 

 

  

 

 

HYPOTHETICAL cuevs’
FOR 140 TONNES OF ,I

 

THORIUM'——\.‘

 

 

 

 

 

 

_ESTIMATED LOCUS OF
T MAXIMUM PERFORMANCE.-«:_

 

 

 

 

 

 

 

 

 

 

 

 

FUEL Y'ELD“’('%'/yr_') S

wlfh Vurymg Thorlum Inventory
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o

 

 

 

 

 

 

 

 

 
 FUEL CYCLE COST AT 80% PLANT FACTOR (mills/kwhr)

FUEL YIELD AT 80% PLANT FACTOR (% /year)

UNCLASSIFIED

. ... ORNL-LR-DWG 58854

 

- 9L -

 
(L

- T =

Several calculations for a representative opersting condition were made
using the GNU and ERC-5 programs for the IEM-TO4 to determine the effect of
these 1 variations. The plots of Figs. 6.7 and 6.8 show the ef:fi'ect for cases
in which fission product resonence absorptions were included -and for cases con-
sldering fission products to be ‘l/v gbsorbers. There 1s considera‘ble veriation
in the fuel yleld, as measured by the horizontel distance between corresponding
points on the curves, between the f'recommended ‘eurve and the high and low epi-
thermal eta curves. 'J.‘he deviation from the recommended‘ relue is ¥ 2.5 to

% 3%/yr in fuel yield. 1In fact, using the more pessimistic values of 1 makes
it difficult to attain fuel yields of as mnch as h%/yr even at fuel cycle costs
a¢ high as 1.6 mills/kwhr.

On the other hand the choice of high or low epithermal n-233 does not have
& strong influence on the fuel cycle cost. This effect 1s measured by the verti-
cal difference between corresponding ;points on the three curvee. This difference
is epproximately t 0.06 mills/l:whr from the recomended curve.

The various velues of 1-233 are tebulsted in Table 6.6.. The column headed
n(MIR) contains values from experiments performed at the Materials Testing Reactor
end are the velues recommended by Nestoru for éuse in this study. '.l'he coluins
headed n(+ 10%) and n{- 10%) contein the extreme velues used to obtain the two
curves of Figs. 6. T end 6.8 for comparison with the recommended values. Values
in all energy groups differ by 10% from the MTR values except in group 31 where -
the difference is only 5% end in groups 32 through 3k where the velues were not
changed. The other values headed n(RPI) are presented for comparison since these
are recent data from & study by.Y_ea.ter.lg The RFI set of velues are not thought
to be more relieble than the MTR velues; however, in the energy renge 30 ev -

1 kev the RPI date represent the only measurements thet have been made.

6.6 Effect of Value of Pa.-23; Resonance Integra.l on MSER Performence

‘A second miclear constant that is perhe.ps not known very accurately is the
value of the resonance integra.l of Pa-233. Figures of 600, 900, end. 1200 barns
have been mentioned for the value of this integra.l. As mentioned in Section 3.7,
& velue of 1200 ba.rns wa.s chosen for this study.f Hovever, in order to determine
vhet efrect & lower value of the resonance integra.l would ‘have on reactor per=
formance s several ca.lcule.tione were made at repreeenta.tive opera.ting conditions
nsing & 900-barn resona.nce integra.l for comparison vith the 1200-bern cases., The
resulting curves for the reactor performance are plotted in Fig. 6.7 end 6.8.
 

Grdmp

o

W 00 =~ o8 F W N

o e e B M R
colbakR&BES

S RBRBES3

W oW W W o NN
RUrEsIRNRER

1x20° -2 x10
3x20° -1 x10

. -18 -

. Table 6.6. Group-Averaged Eta Velues of U-233

| Efiergy‘(ef)
L x l06
2x10% -1 x 10
6
6

1 x10° - 3 x 10°

3x10" -1x10°
1x10% -3 X 10" \

3_x1103 -1 x_lou

1x100:23x10°
ko -0

150 - k0o
100 - 150
90 - 100

80 - 9 .
65 - 80

33 - 37
30 - 33

25 - 30 -

20 - 25
17 - 20
13,5 = 17
10 - 13.5

7.5 - 10

55 - T.5

o35
g iy

1k - 2.5 -

0.8 - 1.
therma.l - 0.6

- thefmal

-1 fx_‘.mT -
6 .

2,025

- 3.051

. 2.583'
2.h21

- 2.259
2.133
2.052

L 2.025 .

2,016

 

Lo

1.944
1.953
1.933

2,052

1.863
1.962

S
1Ny

2.2
'2.2’_9_‘__ 3
2,28

2.24

a2

 

2.16

' 2.16

2.16

; 2.17 "
2.28

2.07
2.18
2.05
1.91

2.23

2.29

2.28

n(-108) - w(OmR)  n(+10%)
| o 3.39
- 2.87
- 2.69
B
2437
- 2.8
225

3.729
- 2.959
2,760
 2.60T
2,508
©2.475
2.6k
20475

 

2,376
23716

2.376

2.387
2387
oo
- 2.39%8

2.255
2200
2.3k
. 2.29
- 2.28
- 2.28

_a(eem)
3.39

287
- 2.69

2.51
2.37

- 2.25 .

2.24
2.25

1.9
T2
1.78

1.54%
1.68

- 193
10.96 -

1.97
2.06

1.91
1.93
R

1.88
- 2.06

1.91

1 1.96

10%

1.96

2,07
2.23

2,29

| 2,28

2,28

G
A

[ ]

»

-79 -

. The improvement that the lower valne of the resonance integral makes in -
tie MSER performance is hardly epprecisble. The fuel yiel&,-measured=by the
horizontal  difference between corresponding points, is increased by perhaps -

0. 25%/&ear, the fuel cycle cost, measured by the vertical difference, is lowered
by ebout 0. 01 mills/kuhr The small effect on perfbrmnnce is nnderstandable _
vhen it is considered that losses to Pa-233 account for only about 0. 5% of the

' neutrons born per fuel absorption using the 1200-barn resonance integral., How-

ever of all the neutron absorptions in Pa-233, approximately 80% occur at epi-
thermal energies. | N

The_comparative 900—barnkresonance.integral calculations were performed
using the ERC-5 code with adjusted values of the reaction rate coefficients,
vhich ere defined above in Section 4.k. The adjustment was made using output
deta from & Cornpone finite reactor calculation giving the absorptions in Pa-233
as a fhnction of energy. The reaction rate ooefficients were calculated sepa-
retely for the thermal absorptions end the epithermal absorptions, the epithermal
value being decreased by the ratio of the resonance integrals, i. e., 900:1200.
The two values for the coefficients were summed to ‘obtain the total reaction rate
coefficient as shown by Eq 1k, - ‘ o -

[ s A(Pa.) | A(Pa) S
th
C(re) 1n core = [ N(Fe) (1200 )] @)
_ - . core |
where
C(Pa) = reaction rate coefficienf;of Pa-233
A(Pa)th = neutrons ebsorbed by Pa-233 et thermsl energy
- per neutron born in core
A(Pa)épi = ‘neutrons ‘ebsorbed by Pa-233 at epithermal energies

~ per neutron born in core
N(Pe) = homogenized atomic concentration Pa-233 in core, atoms/cm

A similar expression-was used to calculateic(Pa) for the blenket, and these
adjusted coefficients were used in the equilibrium reactor calculations.

6.7 Effect on MSER Performance of Adding ZrFs to Fuel Salt

Recent developments in fuel technology for the Molten Selt Reactor Experiment'
(MSRE) have indicated that fuel stebility is enhanced by the addition of nominal
amounts of ZrF¢ to the fuel selt. Zirconium acts es & "getter" for oxygen
 

 

- 80 -

preventing the union of oxygen and uranium.which-results‘in the precipitation -
of'uranium’oxides. The inclusion of zirconium, however, adds an edditionel

neutron poison ‘o the system.

The effect on breeding ratio and fuel yield of ‘adding 5 mole % ZrF@ to the
fuel salt vas calculated for the ranges of values representative of the MSER. |
The results are pdotted in Fig. 6.9. The curves BhOW'the per cent decrease in
breeding ratio and fuel yield resulting from the addition of Zer as & function
of these quantities for a reactor containing no Zng. As might be expected, 7
the detrimental effect of the ZrF4 is more pronounced for the reectors that have o
initially poor breeding performance.,pIn fact the steepness of the fuel yield
curve st fuel yielde of the order of 2-3%/yr suggests that adding ZrFs to & low
performence resctor can Just sbout destroy its breeding potentisl. -

The Zr-containing salt used in the calculations had the composition of fuel
solution proposed for the MSRE: 70—23 5el-l mole % LiF-BeEg-ZrF4-ThF44UFz.

In order to determine the effect of 5 mole % ZrFx on & representative per-,'
formance curve,_the resulte of Fig. 6 9 were applied to equilibrium reector cal-
_ culations for a fertile stream cycle time of 50 days. The results were optimized

according to fuel stream cycle time and fuel salt discard time. The horizontal
difference between corresponding points in Fig. 6.10 shows that Zr decreases the
fuel yleld ebout 0.5%/yr; whereas, the fuel cycle cOst;'measured'by the vertical
| difference, is almost negligibly'affected. The effect on fuel cycle cost is
small because the effect shows up through the loss in breeding credit which is
not a large portion of the total fuel cyole cost.

O
_..8__1_..

UNCLASSIFIED
ORNL«LR~DWG 58655

 

PERCENT DECREASE IN FUEL YIELD
 

UNCLASSIFIED
ORNL-LR-DWG 58656

82 -

 
 

[ 1

- 83 .

7.0 CONCBUSIGNS;;

' The molten salt reactor offers considersble promise as a breeder in the
ThJU'cycle. The prihcipal advantage of this systenm over other breeding systems,
vhich use thorium in the form of the oxide or metel, is in the simplified chemical -
processing*mcthod. The molten salt system is able to use the relatively simple
fiuoride volatility process plus the HF dissolution process for uranium recovery
end dccontamination; vhereas, breeders vhich employ ThOz or thorium metel are,
in the light of current technclogy, resigned to the more complicated and ex-

pensive Thorex process. .

‘The MSER is capable of excess fuel yields up to T%/year vhen operating 80%
of the time. At this high yield, the fuel cycle cost is ebout 1.65 mills/kvhr.

At lower fuel ylelds the fuel cyclercost.is;considerably-imprcved,‘dropping to

perhaps 0a65:mills/kwhr;fbr fuel yields.of'l-afl/year. - However, yields as low

‘a5 this constitute marginal'qperation beceuse uncertainties in nuclear data

1ntroduce uncertainties of about a per cent in fuel yield into the calculetions. -

At & fuel yield of 4%/yeer, the roint at which the income from sales Just balances.

the annusl charge on fissionable 1nventory, the fuel cycle ccst 1s approximately
0.9 mills/kmhr

The largest contribution to thc fuel cycle cost ie made by the fuel stream
processing plant which acccunts for ebout 414 of the cost at the high processing
retes (high fuel yield) and -ebout 30% of the cost at the low processing rates.
Another item that makes & maJor contribution 4o the cost et the high processing
rates 15 the fuel salt discard, acccunting fcr slightly more than 20% of the

tctal, however at the low processing rates, salt discard acccunts for only about L

b, o o - o o o
The fertile stream processing plent cost emounts to only 12-15% of the total.

, Thorium inventory for the 6000 ft3 fertile stream emounts to 8174 of the cost. -

Since the . thorium inventory is constant (270 tonnes), its cost is & larger por=-
tion of the cost fcr those cases that have the most favorable fuel cycle costs,
The same is true for the thorium carrier which accounts for 10-22% of the fuel
cycle cost over the range of fuel yields from T4/yr to 1%/yr. |
 

 

'._;{84 -

Fissionable inventory (including PE-233) is only slightly affected by the -~/
processing rate. of the fuel stream over most of the range of poison fractions -
studied.  However, at the: fast fuel stream cycle times the holdup 'in the chemical
processing plant begins to contribute’importentlj to the inventory. The principal
fector increasing the inventory is the fertile stream cycle time. In going from
& fertile stream cycle time:of 35 deys to,aoofldeys,,the fissiOnEble inventory
increases from about 8&0 to 1280 kg. At the same time, fertile stream fissions -
increase from 1.3% to 6. 6%>of the totel fissions. The fissionable inventory -
accounts for ebout 15% of the ‘fuel cycle cost at a fuel yield of 1. 5%/&r and for
about 4% et & yield of 7%/yr. Lo '

The breeding performance of the molten salt reector is especially sensitive
to the value assigned to the epithermel.fission cross section of U-233 since:
ebout 30% of the fissions occurs at epithermel»energies;’ E@uilibrium reactor
calculations for a representative set of operating conditions indicate that the
fuel yield may very s much es ¥ 2.5 to ¥ 3¢/yr for varistions of £ 10% in the
velue of the epithermal cross sections of U-233 from- the set used in these
ca.lculations. 3 ' ' ' e

The inclusion of 5 mole % ZrFq in the fuel salt to emhence stebility de-
creases the fuel yield about 0.5%/year; however, the concurrent fuel cycle cost
is negligibly increased. The MSER already suffers from.having relatively high
neutron ebsorbers in the molten salt carriers, as compared with graphite and
heevy water in other ‘breeder reactors, and the eddition of any other atom with
appreciable cross section can only lower the breeding performance. ' o

There ere two weys of improving the breeding performence of the MSBR. These
are (l) determining the optimum C:U ratio and (2) increasing the thorium.inventory
in the blenket. In regard to the C:U ratio, it is believed that the value of
approximately 5100 used in these calculetions is near the optimum end that only
a very slight‘improvenent might be expected'by:changing‘the‘reector composition.~
 The most significant improvement in the MSBR breeding performance ‘can be made -
by increasing the thorium inventory in the fertile stream. In the blanket,

Pa-233 competesvwith thorium for neutrons; hence the losses to Pa<=233 are in-
‘rersely proportional to the-thorium concentretion. However this‘improved breed- |
ing performance comes at the expense of additional'charges for thorium and fertile i:;
 

ay

"

-85 .

salt inventory, end the net effect On'therfuel cycle cost will be en increase. -
Above & 270-tonne thorium inventory, vhich was used in this study, the increased
breeding credit is insufficient to offset 1nsreased thorium and fertile salt ,

inventory charges.

The molten salt reactor concelved for,this.stsdy necssssrily includes
some elements of design which are perhsps beyond current technology, e “.g., leak=
proof graphite-to-mstal Joints end impervious graphite that permits minimum xenon
dbsorption.' In chemical processing, further demonstration of the fluoride vola-
tility process and the HF dissolution process is necessary to suyply adequate
design information. e - |
 

12.
13.

1k,

15.

| -8._0 REFERENCES

L. G. Alexander, et aJ.. ’ Thor:l.um Breeder Reactor Eva.lus.tion. Part I,

Fuel Yields and Fuel Cycle Costs in Five Thermal Breeders, 0RNL—CF-61-3-9,

“R. C. Robertson, Sizes of U. S. Steam Electric Plants, ORNL CF-59-5-130,
}May26 1959 - e . .

~'W. D. Burch, D. O, Campbell, and H. O. Weeren, Processing Me'bhod.s, Fission

Product Poisoning, Fuel Cycle COSts for Fluid Fuel Reactors y ORN'L CF-60-1+-1.

T April 1960

| 'C. W. Nestor, l»fizltigroup Neutron Cross Sections, ORNL CF-60-3-35, March 15,

1960.

J. E. Evens and R. O. Fluharty, "Eva.luation of Low~Energy Cross-Section
Data for U-233," Nuc. Sci. Eng. 8, 66 (1960).

R. W. Stoughton a.nd J. Halperin, "A Review of Cross-Sections of Par‘bicular
Interest to Thermal Reactor Operetion, Ruc. Sci. Eng. 6, 100 (1959).

E. A. Nephew, Thermal and Resonance Absorption Cross=Sections of the U-233,
U-235, and Pu.-239 Fission Product.s, ORNL-2869, Jan. 18, 1900.

C. L. Davis, J. M. Bookton, and B. E. Smi'bh, A Multigr oup, One-Dimension
Diffusion Progrem for the IBM-?Oh , GMR-101, Nov. 12, 1957

V. E. Kinney, COrnpone - A Mtfl.tigroup, Mtiregion Reactor Code, ORNL-2789,
in prepa.ra.‘bion. ' o

 

L. G. Alexender , ERC—Q Progrsm for Comggt% the Equilibrium States of Two-
Region, Thorium Breeder Reactors » ORNL CF-60-10-07 ) Oct. 20, 1960.

Weinrich and Associa.tes, Process Designs end Estimated Costs of Chemical

- Plants for Processing Molten Selt Reactor Fuels, & report to the Chemical

Technology Division of the Oak Ridge Na.tional Leboratory, June 1959.

H. G. MacPherson, et al., Interim Report on Fluid~-Fuel 'Ihermal Breeder
Reactors, ORNL CF-60-3-31 (Revised), Merch 15, 1960

H. G. MacFherson, Molten Selt Breeder Reactors, OR'NL CF-59-12-6h (Revised)
Jan. 12, 1960.

'I. Spiewek and L. F. Parsly, BEvaluetion of Externsl Holdup of Circulating
- Fuel Thermal Breeders &as Related to Cost and Feasibility, ORNL CF=60-5-93,

Hay 12, 1960.

H. R. Payne and J. C. Moyers, Determination of Capital Ccsts of Steam Cycle i

Fquipment and Over-all Plant E:E‘ficiency for Three Breeder Res.c'bors une

 

published date.
[ ]

o'

‘g¢b

- 87 -

L. G. Alexender, Oak Ridge Nationel Leboratory, unpublished deta.

L. Dresner, Tebles for Computing Effective Resonance Integrels, Includi
Doppler Broadening of'Nuclea.r Resonances, ORNL CF-5_5-9-7 Sept. 19, 1955).
C. S. Welker, Reactor Controls, ORNL CF-57-1-1 (Jen. 5, 1957).

M. L. Yeater, R. W. Hockenbury end R. R. Fullwood, Ete of U-233 from 1l ev
to 800 ev, Rensseleaer Polytechnic Institute Report , June 1960.

L. G. Alexander, et al., Thorium Breeder Rea.ctor Evaluation. Part-I.

- Fuel Yields end Fuel Cycle Coste in Five Thermal Breedere, ORNL CF-6l-3-9

(Appendices, Part 1), March 1, 1961,

- Je W. Miller, Evaluastion of & Deuterium-Modersted Gas-Cooled Breeder

Rea.ctor, ORNL CF-61-3-2, March 1, 1961.
 

9.0 APPENDIX
o

«k ) ?

Te.ble 6 2. Performance of a Molten Salt Breeder Reactor for Several Values -

-of Fission Product Poison Fraction.

Fission Product Resonance

Absorptions Included :Ln Ca.lculations.

 

Fertile stream cycle time (days) 35
- Case No. X 2 3
' Poison fraction 0.02 0.04 . 0.06
Volume fraction fertile stream in core R dTl 0.068 0.066
Volume fraction graphite in core ' 0.T69 0.772 0.7TT5
Carbon: Uranium ratio in core. 5020 5040 5050
Fuel stream cycle time (days) C12 o5 84
' Fuel salt discard cycle time (days) 145 - 430 1550
Fraction of fuel stream sold as product . 0.0035 ~0.0L04 - 0.009%
Fraction of fertile stream sold as product 0.0182 0.013%  0.00T4
- Fraction of fission i.n fertile stresm : 0.0139 0.01.33 . . 0.0126
| -2&“” L Lo D
‘Fuel stream composition (a.toms/cm )(10 _ R Co e S -
. U-233 . , 0.8892E-4 0.8855E-k . 0.8812E-4
U-23h | 0.2516E-4 0.26158-4 0.2729E-4
U-235 - ' ~ 0.8Th9E-5 0.9266E-5 0.9863E-5
U-236 - 0.1069E-4 . 0.1278E-4 0.1575E=4
Np-237 o . 0,6001E-T 0.2B4TE-6 0.54TOE-6
F.i.BBiU.I?(a) . o ; ey
Xe=135 "~ 0.1260E-9 .12703-9 - . 0AdZTLE-9-
Carrier(c) 0.208581 0 2085E-1. - 0.2085E-1
Fertile stream comosition (atoms/ )(10'2’*) - L | ' S e
. Th | - 0.4OL2E-2 | .uoz.an-a | - 0.4012E-2
Pa.-233 . 0.644E5 0.1601E-5 ~ 0.1554E=5
U=233 0.1439E-5 0.1L401LE-5 0,1360E-5
U~234 - 0.063E=T  QJA0L3E-T S 0.9590E-8
- U=-235 0. 344 E-LO 0.3444E-10 - 0.3444E=-10
Pisatum(a) - © 0.1892E-5 ~ 0.1808E-5 . 0.1T18E-5
Sm-15L - 0.3716E-9 0.3631E-9 0.3539E-9
" Sm=-149 @ "~ 0.1063E=9 . 0.,1041E-9 0.10L6E-9
Carrier‘’’ . 0.4012E-2 ~ DJMOL2E-2 0.4012E-2.
Note: See end of Table 6.3 for footnotes.

B

 

75
L 5 6
0.02 0.04 0.06
0.0704 0.0679 0.0653
0.770 0.772 0.TT5
5055 5065 5075
12 50 78
1k5 460 1700
©.0.0031 0.0089 - 0.0075
, 0.02k5 = 0.0179 0.0097
. 0.0269 0.0257 0.0244
. 0.8T48E-4 0.8721E-4 0.8689E-4
0.2554E-4 0.264TE=4 0.27T5LE-4
" 0.80UGE-5 0.9431E=5 0.99TTE=5
0.1156E=4 0.1354E-4 0.163TE=k
0.6431E-T 0.2961E-6 0. 5332E-6
- 0.1270E-9 0.1270E-9 0.1271E-9
0.2085E-1 - 0.2085E~1 .'o 2085E-1
0.4012E-2 0.4Q12E=2 . 0. l+01213-2
0.1639E-5 0.1596E=-5 0.1550E-5
‘0.3032E=5 0.2955E=5 0.28T1E-5
0.2711E-T 0.2584E-T - 0.2U51E-T
0.34L4E-10 0.34L4E-10 0.3444E-10
0.3658E-5 0.349LE=5 0.3315E=5
0.T204E-9 =~ 0.TO31E-9 0.68L4E-9
0.2062E=9 . 0.2016E-9 0.1966E~9
0 .11-01213-2 0 . 1'-012E'2

0,4012E-2

'
@
" \O

'

 
Ca.se 'Io._, L ‘. - | S -'1 e L 3 . 1:»

‘ f':‘ Neutrons absor'bed. by J.isted. element o

Table 6.2 - cont'd

 

Per neutron absorbed - in fuel ' L RS e T i e s
Th o - : 0,997 =~ 0.9703 - 0.,9%10 .+~ - = 0.9936 v .9387
- Ta fissions : , ‘ - 0.0019 o080 . 0,008 - 70,0019 v 8. 70,0018
Pa-233 x 2 o019 .~ o.0213 - 0.0107 -  o.0118 o 0113, -« . 0.0L0T -
U-233. - - 0.9168 - 0.9119 - . 0.9063 - 0.9L49 - 00,9013 1 0. 9052
U-235 . - o 0.0832 0.0881 . T0.0937 0 0 - 0.088L 0 040897 0 . 0. 0948
U236 S 0.0108 0.0129" S 0.0188 -7 0 0,015 o v 0.01360 - 0.006k
. Np-237,. Lo | 0.0002 - 0.0011 S 0.002L° . 0.0003 .. 00.0012° 0 s O, 0021"
L Xe-l35( ) ‘ - .0.0050 0.0050 J0.0050 - 7 0.0050. ... 0,0050 . . 0.0050 -
- Sm-151 + Sm-lk9 | 0.0001 = 0.0001 : 040002 T 0.0002 S0 0,0002
. Fissium = - 0.0207 -~ . . 0.0407 T0.020h L T o082 0. 0613
- Carbon -, : 0.0286 0 0.028T7. 0.0286 . ..  0.0287 -~ - 0.0288
Fuel_c_arr:l;?a(i;' . , 0.0302  0.0302 0.0302 - . 0.0302 . 0.030L
_ , , o |
0

 

e ..

Th carrier‘ ™ | 0200 -~ 0.0196 20402000 7040200 . 0.0200
Corrosion products - 0.0008 - 0.0008 00,0008 U 040008 L 0.0008
‘Delayed neutrons 0.0043 - 0.0043 - . 00043 i _'0.00h3" e 0.0083 T
Leakage o | : ' . 0.006 - 0.0016 - ,:Nommefirpfoom&fifigf}omm'f
Fuel processing L C.0.0022 0.0022 - L 0 0022 ' 0.0022. . .0, 0022 . -

5aRaE

Togobgoboo
o

.21.36_.1- S 2118" U a0

Neutrons born per fuel absorption (qe) - 2.2136 2222 2.2 bl L o _ S
Sl loso o loem

Net breeding ra.tio o L ' -1.0753 ' o l-051-9‘.i' v

. "‘l‘-".
5

 
“y " . . - # ' . | ) | ) - .? ’

~ Table 6.2 - cont'd

- Inventory per station (kg) . : S | _ o o _ | .

| T2 in fertile streem . = . . - 263,000 . 263,000 263,000 263,000.'- . 263,000 263,000
Th in processing 0 T500 . 1500 - 1500 3500 3500 - 3300

Total Thorium L - 270,500 270,500 270,500 266 500 266,500 266,500

© . U=233 in fertile stream . . - - - . - 946 92.2 89.5 199 . . 194 - 189 -
© Pa-233 in fertile stream = - 108 - - 105 02 108 \ 105 _ 102
U-233 in fertile stream processing* . 2.7 - 2.6¢ ' 2.6 . 2 2.6 2.5
" U-233 in fuel stresm, in.reactor 196 195 - 195 . 193 o192 _ 192
U=235 in mélistreamyin.-reactor(e) < 18.5 20,6 21.9 19.9 aa 22,2
‘U-233 in external fuel circuit eth- .. . 213 ‘ L . 269 268 267
U-235 in external fuel circuit = 27.2 - 28.8 - -30.6 - 27.8. 29.3 31.0
. U=233 in fuel processing : . T 83,1 10.1 6.1 ‘ k2. : 10.1 6.5
- U-235 in fuel processing o kL3 ; S la 0.7 L.y 1.1 0.7
- U-233-in fuel reserve - 8. . 67.7 - . 66.9 _ 67.0 66.4 ' 65. 8 3
. U233 + U=235 in fuel dump tenks S 5,67 . 5.8 51.0 5.1 : 51.2
~ Total fissionable :anentory o 896 85&.9: | 845.6 , 991.1 %78 937.5

 Fuel salt (mel excluded) = 31,500 30,000 29, 500 - 29,700 - 29,430
 Hlanket; salt (mF, excluded) | - 201,000 201,000 201,000 - 198,000 198,000 198,000

Doubling time (full power- years) L 11‘.7 . 162 1303 13.3 18.3 35.5
Fuel yield at 80% plant factor (%/yea.r) o .8 k9 2.6 6.0 bk 2.3

‘-‘r6'-'-
~ Table 6.2 - cont'd *

Fuel cycle cost (mills/kwhr) | o | B e S LT e F
' Uranium inventory o ' 0.076 0.073 0.07T2 .~ 0.085 - 0.081 0.080
- Thorium inventory o 0.132 0.132 - 0J320 0 . 0,30 1 0.130 - 0.130
Fuel salt inventory . ' - 0.023 0.022 0.022 0.023 = 0.022 . 04022
Fertile salt inventory L 0.166 0,166 0,166 0.164 S 0,16k - 0,164
. Fuel processing plant - : 0.654 0.265 ' - 0.207 - 0.650 - 0.269 . © o 0.207
- Blanket processing p%ast . 0.261 0.261 0.26L 0.6k 0.164 0.6k
" Thorium amortization . -0:030 0,030 0.030 . ©0.030 - - 0.030 .- 0,030 -

Fuel salt replacement 0.456 -0.145 - 0.040 . - 0.455 0135 - 0,036

. Fertile salt replacemeni-?&) | 0.034 0.03h 0,034 0.3 003k oo

‘-Gross fuel cycle cost U . 1.832,- | ‘;128‘ g .96h‘ ; 1,735 _ .‘.l 029 | 0.867
_ Breeding credit . : .0.130 0.090 S 0.04T . 0.127 - ]‘0.088 . ©r0.045
et fuel cycle cost T2 S L8 o 1.608 0.9 0.822

~ Processing rates 3 S S ' ‘ R el o
Spent fuel 5& /da.y) L : _uh 2 10'.h : 6.3 Wy.2- . 10.6 S 68

- -Thorium (kg/day) R T500 7500 700 '3%0 350 . - 3500
Thorium replacement (kg/day) S - 39.T - - 39.6 39.6 C39. . ' 39.1 0 .. °39.0
Fuel salt replacement (kg/day) aT 68.9 ©18.90 ooars 0 B TN
Excess fissile atoms produced (kg/day) 0 .208 Ok . 0.076 ©  0.203 S0 .lhl , - 0.072

Fertile strean loa.d.ing, (gm U—233 + Pa.-233)/kg Th - ]_o.'n .0.75 . 073 "_".1 . _*J_,-.a = Co1a o1

- 86 -

 
-k ?

 Table 6.3.

' Fertile stream cycle t:l.me (days)
Case No.
Polison fraction '

Volume fraction fertile stream in core
Volume fraction graphite in core
Carbon: Uranium ratio in core

Fuel streem cycle time (days)

‘Fuel salt discard cycle time (days)

- Fraction of fuel stream sold as product

- Fraction of fertile stream sald as product'
. Fra.ction of fissicms in fert::l.le stream

o Fuel stream composition (atoms/cm
' U-233 .
U-234
U=-235 -
U-236
Np-237
Fissi (a)
Xe-135

‘ Ca.rrier( )

5o

' Fertile streanm composition (atoms/ )(10'21"

Th .
Pa-233 -
U-233 -
U=-234 - _
. Feopm(®
~: Mssiwm
Sm-l'?.
Sn-149
Ca.rrier(d)

3)(10'2‘* (a ): |

)f'

 

- ’

Performance of a Molten Salt Breeder Reactor for Several Values

of Fission Product Poison Fraction.

Fission Product

Resonance -Absorptions Included in Calculations.

 

100 .
T 8 9
0.02 0.04 0.06
- 0.0TOW 0.0678 0.0652
- 0.770 - - 0.TT2 0.T75
. 5070 - 5090 2090
12 56 86
150 - hoo 1500 -
- 0.0029 0.009L 0.00Tk
0.0348 S 0.0331 . 0.0315
- 0.8663E-4 0.8639B-4 - 0.8614E-4
.. 0.25T5E=4 0.266TE=L 0.2763E-4
 0.1190E~k - 0.1403E-4 0.16T5E-4
~ 0.66TTE-T 0.3402E-6 .0.5936E~6
. 0.,1270E=9 - 0.127T1E-9 .. 0.1271E-9
~ 0.2085E-1 ' 0.,20858-1 0.20853-1
0.4012E-2 = 0.4012E-2 0. hmam-a
0.1636E-5  0.1592E-5 0.1548E-5
O.4OQLE-5 0.3898E-5  0.3792E-5 -
0.3965E-T 0.3TT6E=T 0.3586E-T -
0.3444E-10 - 0.3444E-10 0.3444E-10
0.4728E-5 0.4506E-5" - 0.4284E-5
0.9327TE-9 0.9096E~9 0.88%9E-9

0.4012E-2 -

 

0.4012E-2

0.4012E-2

o

200

 

0.4012E-2

0 ohOlaE-a

!
w

10 11 12
0.02 0.04 0.06
0.0698 0.0673 0.0647
0.770 . 0.773 - - 0.TT5
5140 250 260

- 1.8 50 . 8y
150 L5 - 1500
0.0022 0.0062 - 0.0052
040370 0.0259 0.0132
0.8346E-4 0.8340E-4 0.8334E-4
0.2634E-4 0.2TL4E-k 0+2T95E-L

. 0.9368E-5 0.9781E~5 0.1021E-4
0.1327E=4 0.1531E-4 0.17T9E=4 -
0.T327E=-T 0.3349E-6 0.6178E-6
0.1270E-9 0.1271E-9 . = 0.1272E-9 ,
0.2085E-1 0.2085E-1 = 0.2085E-1.

. 0.40L2E-2 0.4Q12E-2 0.4012E-2
0.1626E-5 0.1584E-5 0.1541E-5
0.TTO3E=5 0.7515E-5 0.7320E-5
0.1061E-6 0.1012E-6 0.9623E-7
0.3444E-10 0.34l4hE-10 0.3444E-10
0.8794E-5 0.8390E-5 0.7986E-5
0.1T43E-8 0.170LE-8 0.1658E-8
0.4993E-9 0.4879E~9 © 0.4763E-9

0.4012E-2

 
 

Table 6.3. - cont'd

- Case No..."'}-' - - o | T 8 o 9 - 10 | ll

Neutrons absorbed by listed element = -

' per neutron absorbed in fuel
. Th ‘

. Th fissions
. Pa-233 x 2
U-233.

. u-23h

- U-235

- U-236
- Np~237

~ Xe-135(0).
" Sm=151 + Sm-149
Fisslum

- " Carbon

- Fuel ca_rri?r(i)

' Th carrier

Corrosion products
Delayed neutrons

Leakage

e ‘.Fuel processing

_ Neutrons born per fuel absorp'bion (ne) -
e Net breeding ra.tio

0 9915 :
10,0019

0.0118

- 0.9139 -
0.091k
0.0861

0.0120

0.0003.
- 0.0050
~0.0003 .

0.0218
- 0.0286
- 0.0302
0.0200

0.0043 .

0,0016

0.0022

. 2.2128
- l.0T2L

0.9370
0.0018 -
0.0106

T 0.097T9

0.0168

0.0023
' 0.0050

0.0003

'0.061T

0.0288

0'.0301‘_,'- ‘
00,0191
‘ ,_0.000Bj ‘
7 0.0043
. 0.0016 -

2.2101.
| 1.0247

0.9853

0.00L9

0.0116

0.9109
0.0933 |
. 0.0801
- .0.,0133 -
0.0003
10,0050

o 0.023% -
0.0286

. 0.0302° -

- 0.0199

0.0043

- 0.0016

- 0.0022

2.2119
. 1.0682

0.9592 .-
0.0018
- 0.011)
0,907
- 0.0963
. 0.0928
0.0013 -
. 0.0050 -
0.0005

0.0432

0.0287
o 0.0302
-‘00191; -
) ooou3 N

- 0.0026
10,0022

20210‘,?'; )

 
’} .’ ) ‘ ‘,;;. .» e “ g awwn

)

Table 6.3. - cont'd

Case No. S S - . _ T | 8 = 9 10 n ‘ 12
Inventory per station (kg) . - = | - | o R .
Th in fertile stream. - \ 263, 263,000 _263,000 263,000 263,000 . 263,000
Th in processing - : 2600 - - 2600 2600 _ 1300 - 1300 s 1300
Totel thorium . . | . 265,600 265,600 265,600 264,300 264,300 . 26k, 300
U-233 in fertile stream ' 263 256 249 507 Loy 481
Pa-233 in fertile stream 108 , 105 102 ' 107 - o 10
U-233 in fertile stream processing : 2.6 2.6 . -~ 2.5 ‘ 2.5 . 2.5 2.4
U-233 in fuel stream in reactor 191 : 191 ' 130 184 184 184
U-235 in fuel stream in reactor(®) * .  20.1 21.2 223 - 20.8 - 2.8 22,7
U-233 in external fuel circuit 267 266 - 265 251 251 - 251 .
U-235 in external fuel circuit 284 29.6 . n,2 29.0 30.3 1.7
U-233 in fuel processing ‘ k20 8.y 5.8 o ko 9.6 5.8
. U-235 in- fuel processing - : R 1.0 0.7 © kT - 1la 0.7 -
‘U-233 in fuel reserve - o - 66.2 . 65.7 65.1" _ 63.4 : 63.0 62.7 "
U-235 in fuel reserve . - ' - 6.9 " T3 7.6 0 T4 T.7 '
U-233 + U-235 in fuel dump tenks 50.6 50,7 - 50.9. b9y 49.3. 49.5 5
Total. . fissionable 1inventory L 1049.9 1005 o 992.1 1272.7 - 122k.0 1206.2. \n
Fuel - sa.lt (mel excluded) S 31,500 29,500 29, 500 - 31,400 29,600 29,hoo
Blanket salt (Thrh excluded) Lo 197,340 197,340 197,340 196,400 196,400 196,400
Doubling time (full pover years) - 14.3 20.3 ' 39.5 ~18.3 26.6 - 55,0

Fuel yleld at 80% plant factor (%/year) - - 5.6 3.9 2.0 b . 3.0 _ 1.5

 
Table 6.3. - cont'd

Case No. _ S . 7 8 9 _ 10 - - 1 o 12
Fuel cycle cost (mills/kwhr) o S T e o R e
' Uranium inventory , o 0.090 o "0,086 : 0.085 - 0.109 . 0.105 0.103 .
.- Thorium inventory . ' 04130 0.130 0.130 o '0.129 o 0.129 0.129
~Fuel salt inventory R S 0.023 . " 0.022 0,022 -0.023 0 © 0,022 . 0.021
_ PFertile salt inventory 04163 . 0.163 ' 0.163 - 0.162 - .. 0.162. . 0.162
Fuel processing plant : 0.641 0.250 © 04207 - . 0.653 0.269" ‘ 0.207
Blanket processing p%ast S ' 0.139 - 039 0139 0 0.092  0.092 . 0.092
- Thorium amortization - 0.030 ' 0.030 - 0.030 .- 0.029 0.029 - C.029
Fuel salt replacement ( )' : 04h0 - _ 0.155 L 0,082 039 0.140 0.0kl
Fertile salt replacement 'S . . 0.033 ‘ 0.033 0.3 . 0.033 ~0.033 - 0,033
Gross . fuel cycle cost : o 1.689 - 1.008 0.8 . - 1. 669 | o 0.981 0.817
Breeding credit - : e e - 0.125 © 0.084 , 0.043 0.8 _ 0.078 o T 0.037
Net fuel cycle cost : - - L.564 - o.92!+ 0. 808 - 1.581 - 0.903 . 0.780
Processing raxes : L : | o o T .
| Spent. fuel f‘b3/da.y) S uh,2 9.5 6.2 B9 10,6 6.3
Thorium (kg/day) o - 2600 2600 . - 2600 1300 ‘1300 ~ 1300
Thorium replacement (kg/day) 38,9 38.9 - 38.9 -+ 38.8 ‘ 38.7 - - 38.7
- Fuel salt replacement (kg/day) . . 210 \ Th.0 19.6 - 209 | 66.6 = 19.6
Excess fissile atoms produced (kg/day) 0.199 0.135 o 0.068 0.189 L0425 0.060 '
_Fertile stream loading, (gm U-233 + Pa-233)/kg Ta 1.k 1.k 1.3 2.3 . . 2.3 E 2.2

- % -

 
<)

-

(2)

Footnotes for Tables 6.2 and 6.3

The element fissium is a conglomeration of fission products. A

“fictitious reaction rete coefficient end concentration were as-
- signed to fuel stream fissium as explained above in Section 4. 5

" to -achieve the desired poison fraction. The concentration of |
" fertile stream fissium was calculated by ERC=5 using reaction

(v)

(e)
()

()
(£)

(g)
)

@

rete coefficients developed from GNU and Cornpone output

All Xe-l35 is assigned to the fuel stream. - It hes been ‘assumed -
that & gas purge of fuel end blanket solutions will maintain Xe-135

'absorptions at 0.005 neutrons per neutron ebsorbed in fuel.

Based on Li-T etcms.

*The ebsorption cross section of the fertile stream.carrier was

normalized to the basis of one thorium atom.
U-235 in fErtile stream is negligible.

Includes thorium burned up in breeding plus thorium discarded on

'20-year cycle. |

,FErtile selt is discarded on a 20-year cycle to maintain blanket
_fission products et a tolerable level.:

The concentrations are written with the letter "E"‘used to denote

the exponent, e.g., read O. 8892E-k as 0.8892 x lt)"l‘L

iReplacement selt for fuel and fertile stream carrier is assumed
to contain Li that is 0.0l atom % Li-6. Absorptions are based on

o equilibrium.Li-G concentration for this feed. -~

e e o et
 

1-10.
1.
2.

- 13.

17.

18,

| f'20
21.

L 22.
23.

2k,

25.
26.

28.
29.
30.
3i.

33.

34,

35 -
36.

37.

38.
39«

ko.
b1,
ke,

43,

L,
45.
46.
4.
18,
k.

.
15. ..
16,

G. 'Alexé.nder

E. Beall -

8. Bettis S
. Blankenship

L-' BOCh

. G. Bohlmann' : , - o

'aOigcampbelli ‘ '
-~ H., Carr -
L. Carter-
. I. Cathers
H. Chapman

L. Culler

.G+ Delene

K. Ergen

"E. Ferguson
P. Fraas

‘R. Gall.

- E. Goeller

. R. Grimes

P. Harmond

H. Jordan

R. Kasten
W. Kinyon

I. Lundin

N. Lyon

G. MacPherson
D. Manly -

B. Mcl)onald :
F. McDuffie
W. Nestor

F. Parsly

M. Perry

A. Preskitt
Spiewak

A. Swartout
Taboada

Van Winkle

. Distribution

.
52,

0.

| 53eBh
’§55'56‘
. '-:57."58'
- -59-68..

3 -

71'.-'

72-73-
T

5.

76+

78.

19

- 80.
81 82,

. . 83.
- 8498,

.

G. M., Watson
- "A..M. Welnberg .

-C "E. Winters
"Reactor Div. Libreary

Centra.l Research Library

 Document Reference Library

Laboratory Records
ORNL-RC

" EXTERNAL

 

H. W. Behrman, AEC,
Washington

H Brooks, Hervaerd Uni-

versity

_Do Fo cope, AEC-GRO

D. H. Groelsema, AEC,
Washington

J. F. Kaufmann, AEC,

Wa.shington
L. Link, Argonne

J. W. Miller, K-25
'Ro E. Palfl-er, AEC’

. Weashington =

B. E. Prince, AEC,

- Washington

¥W. Robba, Brookhaven

F. P. Self, AEC-ORO

D. C. Thomas, AEC, -
Washington

TISE~AEC

*