ORNL-TM-1851.txt

 

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OAK RIDGE NATIONA LABORAIORY
operated by

UNION CARBIDE CORPORATION
NUCLEAR DIVISION LTI
for the

U.S. ATOMIC ENERGY COMMISSION
ORNL- TM- 1851
(COPY NO. - 208
DATE - June 12, 1967

SUMMARY OF THE OBJECTIVES, THE DESIGN, AND A PROGRAM
OF DEVELOPMENT OF MOLTEN-SALT BREEDER REACTORS

R. B. Briggs CFSTI RRICES :

HC. si-_f?g_”, ,,,._,__é__i

ABSTRACT

Molten-salt thermal breeder reactors are characterized by low
specific inventory, moderate breeding gain with low fuel cycle cost,
and high efficiency for converting heat into electricity. Studles
indicate they should be able to produce electricity in 1000-Mw(e)
stations for about 2.6 mills/kwhr in investor-owned utilities, a

-cost that is as low or lower than projected for advanced converter
reactors or fast breeder power stations. The fuel utilization

characteristics compare favorably with those of fast breeders.

The present status of the breéderztechnology is being demon-~
strated in successful operation of the Molten-Salt Reactor Experiment.
A two-region Molten-Salt Breeder Experiment to demonstrate &ll the

‘basic technology for full-scale breeders is proposed as the next

step in the development. Design and construction of the MSBE would
be accompanied by a program of fuels, materials, fuel reprocessing,
and engineering development. Development, construction, and startup
of the breeder reactor is estimated to take about eight years and
to cost about $125 million.

NOTICE

This document contains information of a preliminary nature and was prepared
primarily for internal use at the Oak Ridge National Laboratory. It is subject
to revision or correction and therefore does not represent a final report. The
information is not to be abstracted, reprinted or otherwise given public dis-
semination without the approval of the ORNL patent branch, Legal and Infor-

mation Control Department.
BE“MHflKflEUHUflEUfiQmmfimfflflflfl??fil
 

 

 

 

 

 

 

- LEGAL NOTICE

. This report was prcpured as an uccount of Govommom sponsored work, Nellhor the Unltod Siatos,
_nor the Commission, nor any person acting on beholf of the Commission: T

A. Makes any warranty or representation, cxpussod or implied, with respect to the accuracy,

completeness, or usefulness of the information centained in this report, or that the use of -
any information, apparatus, method, or proeess disclosed in this report may not infrlnyo. '

privately owned rights; or

B. Assumes ony liobilities with respect to flu use of, or for damages resulting from the use of

_any information, apparatus, methed, or process disclosed in this report,

As used in the above, *‘person acting on behalf of the Commission” incledes any employao or .

contractor of the Commission, or employee of such contractor, to the extent that such employes
or contractor of the Commission, or employse of such contractor prepares, disseminates, or

provides access to, any information pursvant to his employment or centract wnth the Commilllon,-" '

or his omployment with such contractor,

 

 

 

 

 

 

 

 

 

 

 
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CONTENTS |
Why Develop Molten-Salt Breeders ------ memmemecee e s —————————————— 7
Fuel U%ilization Comparison ------ --Q-;----,—--—e--------e-------,--- 9
Growth of Electric Generating Capacity cemmeemceccememem———e———
Nuclear Fuel Resources ---ececccaea cemmemesemedeecm e e e 10
Fuel-Utilization Characteristics of Converter Reactors -=e-mer=- 10
Fuel Resource Requirements with Converter Reactors ————wee——en—- 12
Fuel Utilization Characteristics of Breeder Reactors ---eeee---- 14
Fuel Resource Requirements with Breeder Reactors ----=-c--eee-a- 1k
Cost-of -Power Comparison --3-4;---' ------- ;;..-,;-------; ------------- 19
Capi%al COStE —m-rcccmmecmnrcrnrcrrcrccrccccnra e c e e e —— 19
- Operating. COBES wmmmmmmcm e m e e e m——— 21
Fuel Cycle and Total Power Costs --------- —m—mmcee e e ——————— 21
1000-Mw(e) Molten-SaltfiThermelereeder Power Plant —--a--;e—-e--¢e4e- 22
Reference Plant Design mcenmemmetm——————— o o o e e 0 0 o e 23
Fuel, Blanket, and Coolant Salts ----=ewe-eccccccmoccaaccaaa- 23
: Flowsheet meemememe——n meesmceenmcmcrana- meeesmceseaaa— —mmae 23
" Reactor Design -Q;-;7-;;--é——--'—-'-..'—--ii--'---.--'-_----d-a:'-'--—"--.“a‘uu-s 26
Heat Exchange Systems remeeemcecene- cemmecocesmeencea e cmsma 33
Fuel and Blanket Processing menecreererrnccesce e — e ———————— 30
Capital-Cost Estimates «----eccwwea- R ——mmsem- 36
Reactor Powver Plant --c-cecaceccana- BT G 1<
Fuel Recyc1e Plant "'-"""---------F----"-fl-l—-—---—n-uu-q —————— h.o
Nuclear Performance and Fuel Cycle Analyses ----;-f----;-_ ------- 1
' Analysis Procedures and Basic Assumptions -ee--e-ecececcaceas 143
Nuclear Performence and Fuel-Cycle Cost --------- ——————————— Ls
Power-Production Cost and Fuel Utilization -

- Characteristics. --f-?f-FF’-’?""f’”-‘-?*?éfl—P-E-—--fif-Ef-f-f—- L5
Alternatives to the Reference Design mmmmeeemeemoemeecanmsaaaaan 45
 Modular Designs -w=-==srecccscesenc—a- reemmecmm——ceaan— wmeswa 50
Mixed-Fuel Reactor =eeremrmeccreccacrcccccnccacccccccanncaccana 53
Direct-Contact Cooling with Molten Lead *'7"'ff‘f'f"f ...... 58

Program fer Development of Molten-Salt Thermel Breeder L '
Power Plants ----------—-——--f--—g—-e-— -------- ee-e--e—--g-—ge-s---- 59
: Steps in the Development é---------------------;-;;-------a ----- 59
Present Status. of the Technology -~ MSRE —-4---------—-;-;-fi-.--- 60

~ Advances .in Technology. Required for a High- - Sl o
Performance Thermal Breeder --------------------;-_-5;---_;-__- 1
: Criteria for the Molten-Salt Breeder Experiment -------;,_—-_---.72
Summary of Plems, Schedule, and COBLE ==mmmemm-= meeeccmcccemcccccanae TS
Molten-Salt Breeder Experiment --------------—-----4—, ----------- S

Engineering Test Unit and Fuel Processing Pilot Plant ~eeewme-ee 5

 
 

 

CONTENTS (continued)

- -Development of Components end Systems «-c-cceeccacaccacacccneacea- 78
Instrumentation end Controls Development —wececcacccccccrcocawoa. T8
Materials Development. -;--------fi-------------—------—---~-—--—- T9
Chemical Research and Development ~=-eec-ccecana - ————— remmm=e T
Fuel and Blanket Processing Development ===ememm-mcecmemcccac--o 80
Maintenance Development ----e-e-cnwccce-- cressmenccmrran——-— —————— 80

~ Physics Program ---=-===- sremrssceeccsancmcan- cemmmpeecnncasnees 80
~ Safety Progrem --------=--cc-sc-sceco- semmmmesecesemocmcommonese 8]

Fig.

Fig.

Fig.

- Fig{

Fig.
Fig.

Fig.
Fig.

Fig.

Fig.

Fig.
Fig.

Fig.

- Figo
Fig.

LIST OF FIGURES

1. Fuel Required for Inventory end cflrrent —
Burnup in Converter Reactors ------------—#----4----5------6 13

2. Total Fuel Requirements for Nnclear Power
- Industry Based on Introduction of Breeder

3. Total Fuel Requirements for Nuclear Power Industry |

Based on Introduction of Breeder Reactors in 1986 ~-eecmee-- 18
k., Molten-Salt Bréeder Reactor FlOWflDiagram.--A--f----sfs--;---.25
5. Molten-5alt Breeder Reactor Cell Arrangement, , :

Plan View -------- cecsmcmcsececemsnere e e n———— cnmcam——- 27
6. Mblten-Salt Breeder Reactor Cell. Arrangement, |

EJ.EVfltion ---- _"""""f" ---- .----7"-,--_-------_----_'-‘-_-,--,“----"' 28
T. Reactor Primery Equipment ----- mecmca—- ;;-—f-;----;;,- ----- 29
8. Molten-Salt Breeder Reactor Core Cell ~--meccccecaaccmceaeaas 30
9. Molten-Salt Breeder Reactor Primary Heat -

EXChanger and Pulnp ---— ------- - e W e AR 4 AR A mmm- V--'-'---..--'--;.--- 31"
10. MSBR Core and Blanket Processing Schefme «r-m-eemammema—n-- - 37
11. MSBR Fuel-Recycle Costs as & Function of . i

PI‘OCGSSing Rates "'-“""""------‘---’-0--------{-—-> -------- ‘—‘— ----- ,.[.2
12. Variation of Fuel-Cycle Cost with Fuel Yield :

in MSER end MSBR(Pa) Concepts -—------g ----- cememsmemcc————— 48
13.-Mblten-Sa1t Breeder Reactor Plan of Modular Units ~eme=cre-- '51
1)%. Elevation of Modular Units =eemees-e--mceeccceeeeaeco- —————- 52
15, Mixed-Fuel 1000-Mw(e) BeactorCeil_Eievation ;-----a-ff--;m-57_

'f,ffir. :
 

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W e

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’,B‘C.- .’

LIST OF FIGURES (continued)

Fig. 16. MSRE Flow D16gTam ==ie-=s=snem-mm=mmomnsmm=mmemmcnmmnmmns 61
Fig. 17. General Arrangement of MSRE --es-meecmmmsmnemmemmncememens 62
f‘ig; 18. Re.a..ctor).V‘e:ssel'-;-'----‘-.-------.-;-.---.-----...'. ................. 65
Fig. 19. MSRE Activities - July 196h-necember 1965. --------;_ ........ 66

Fig. 20. MSRE Activities - January 1966-May 1967 cmremrmememmamacane OT

LIST OF TABLES

Table 1. Electric Utility Generating Capacity memeeseecceere——————— 9

Table 5. = Fuel-Utilization Characteristics of Several
Breeder Reactors --------------—-----,-._-------------- ----- - 16

Teble _“6'. A Comparison of Estimated Costs for Breeder and
' and Advanced Converter Reactors Based on Investor-

- Owned Utilities Charges memeecsecccmccmecdcmr e cesc e ————— 20

‘Table T. Estimated Physical_ Properties of MSBR Fuel, |
Blanket, end Coolant Salte =eewwcecca- cmmcdemcm e —————— ol
Table 8. Reactor Design Values -------- —m—e—e————— m=mmmmmcmsmeceeece 31

Teble 9., Preliminary Cost-Estimate Sunnnary for & 1000-Mw(e)
Molten-Salt Breeder Reactor Power Station

| /MSBR(Pa) or MSBE? --_-_-_-_'----'*'-"--"---'",'---------f-"'-".-‘ 38

T_a'ble 10. .'_'Summary of Processing-Plant Capital Costs
. fore lOOO-Mw(e) MSBR -=s-comomocnococo- S — T — ko

' Table 11. Summary of Annusl Operating and Maintensmce
S __-Costs for Fuel Recycle ina 1000-Mw(e) MSBR meme—mer——e e b1
 Teble 12, Economic Ground Rules Used 1in Obtaining Fuel- |

'_ o | ' ' Cycle COStS ---é-i-m--,:-_--f----f----_---p-—--n-.------.'-'_-._-'.'- ------- 11-3
Teble 13, Behavior of Fiseion Products in MSBR Systems ---emmm-mmmm- bl

LEGAL NOTICE

This report was prepared as an account of Governmient sponsored work. Neither the United b
" Btates, nor the Commission, nor aAny person aeting on behalf of the Commisston: . !
A, Makes any warranty or representation, expredsed or implied, with reapect to the accu- |
racy, completeness, or usefulness of the information contained in this report, or that the use .
- of any information, apparatus, method, or process ¢ dsclosed in this report may not infringe
. -privately owned rights; or .
: B. Assumes any liabilities with respect to the use of, or for damages resulting from the
- use of any information, lppunlns method, or procedl disclosed in this report.
. As used fn the above, ‘*person acting on behalf of the Commission’ includes any em-
i ployee or contractor of the Commission, or omplcyée of sich contractor, to the extent that
.uch amployee or contnctor of the comminlon or employee o! mch oontrlctor prepareu i

 

Table 2. U.S, Nuclear Fuel Resources -------------------- ——————— --10
Teble 3. ) Fuel-Use Characteristics of Several Types o

' of Conver‘ter Reactors ---------nfluu-—-—--- ---------- - 11
Table h ~ Partial Effect of Usoa on Cost of Power --:----i-----'---;--- 15

 
 

 

Teble 1k,

stle 15.

Teble "16.

Table 17.
Table 18.

Taeble 19.

Table 20.

Table 21,

Table 22.

Table 23.

Prqposed Schedule for Molten-Salt Breeder ;
- Experiment ~eececmcccccccccccncncmnmiceccnnccnaa- cmmcmcnae 76

- LIST OF TABLES (continued)

Neutron Balences for the MSER(Pa) and the' |

,MSBRIDesign Conditions e cssesae . — e m e —--———- --------h6

Fuel-Cycle Cost for MSBR(Pa) and MSBR Plants ------ ' ——————— h'r_ :

Powver-Production Cost and Fuel-Utilization
Characteristics of the MSER(Pa) and the MSBR

- Plants -7-5----------5----------1---.---3-----_--_---;__,-hg
Désign Values for_Modnlar Plantsf-; ------ ---;--------;;--_5h.
'Fuel-Cyole Costs fromnModular Plants.-f;-----#-----js-—s--56
Some Performance Date fsf MixedéFuei Reactor T L.
Accumilated Operating Experience with MERE —mmmmemecmmmmee Gl

Comparison of Characteristics of Full-Scale | |
and Pj_],o'l; Plant Breeders =«-—ceecccccccame- —————— _----5-'—-7]4,
_ )

Summary of Estimated Costs for Development,

. Construction, end Startup of the Molten- ' ,
Salt Breeder Experiment -eeeeeccccccccccccmcoraccncansonnn T

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up in the development of the fast breeder.

- R.'C. Briant, who directed the ANP project at ORNL. Briant pointed out
- that molten. fluorides are thermodynemically steable’ against reduction by-
- nickel-based structural materials; that, being ionic, they should suffer

~ pressure ‘and being relatively inert in contact with alr, reactors based
‘on them should be safe. The experience-at ORNL with molten salts during

WHY DEVELOPIMOLTEN-SADT BREEDERS?

Nnclear power, based on 1ight-water-moderated converter reactors,
seems to be'an assured commercisl success. This circumstance has placed
upon the Atomic Energy Commission the burden of forestalling any serious
rise in the cost of nuclear power once our country has been fully committed
to this source of energy. It is for this reason that the development of
an economical breeder, at one time viewed as a long-range goal, has emerged
as the central task of the atomic energy enterprise. Moreover, as our
country commits itself more and more heavily to nuclear power, the stake
in developing the breeder rises: breeder development simply must not fail.
All plausible paths to a successful breeder must therefore be examdned
carefully. . ,

To be . successful & breeder mist meet three requirements. First, the
breeder must be technically feasible. Second, the cost of power from the

' breeder must be low; and third, the breeder should utilize fuel so effi-

ciently that a full-fledged energy economy based on the breeder could be
established without using high-cost ores. The molten-salt breeder appears

to meet these criteria as well as, and in some respects better than, any

other reactor system. Moreover, since the technology of molten-salt
breeders hardly overlaps the technology of the solid-fueled fast reactor,
its development provides the world with an alternate path to long-term
cheap nuclear energy that is not affected by any Obstacles that may crop

The molten-salt. breeder, though seeming to be a by-way in reactor
development, in fact represents the culminetion of more than 1T years of
research and development. The incentive to develop a reactor based on
fluid fuels has been strong ever since the early days of the Metallurgical
Laboratory. In 1958 the most prominent fluid fuel projects were the:
liquid bismuth reactor, the aqueous homogeneous reactor; and the molten-
salt reacter.f In 1959 the AEC assembled & task force to evaluate the
three concepts. The principal conclusion of their report! was that the
"molten-salt reactor has the highest probability of achieving technical
feasibility. | : .

 

operation of the Mblten-Salt Reactor Experiment. To thoservho have
followed the molten-salt project closely, this success 1s hardly sur-
prising. The essential technical feasibility of the molten-salt system
is based. on. certain'thermodynamic realities first pointed out by the late

no radiation damage in the liquid- state, end that, having low vapor

the intervening years has confirmed Briant's chemical intuition. Though
some technical uncertainties remain, particularly those connected with
the graphite moderator, the path to a successful molten-salt breeder
appears to be well defined.

AN

 
 

We estimate that a 1000-Mw(e) molten-salt breeder should cost $115
per kilowatt (electric).and that the fuel cycle cost ought to be in the
range of 0.3 to 0.4t mill/kvhr(e). The overall cost of power from a pri-
vately owned, 1000-Mw(e) Molten-Salt Breeder Reactor should come to
around 2,6 mills/kvhr(e). In contrast to the fast breeder, the extremely
low cost of the MSBER fuel cycle hardly depends upon sale of byproduct
. fissile material, Rather, it depends upon certain advances in the chemical
processing of molten fluoride salts that have been demonstrated either in.
pilot plents or lsboratories: ‘fluoride volatility to recover uranium,
~ vacuum distillation to rid the salt of fission products, .and for highest
. performonce, but with somevwhat.less assurance, removal of protactinium‘by
liquid-liquid extraction or ebsorption.- S

The molten-salt ‘breeder, operating in the thermal Th—-saU cycle, is
characterized by & low breeding retio: the maximum breeding ratio con-
sistent with low fuel cycle costs is estimated to be ebout 1.07. This
. low breeding ratio is compensated by the low specific inventory* of the .
MSBR. Wherezss the specific inventory of the fast reactor ranges between
2.5 to 5 kg/Mw(e), the specific inventory of the molten-salt breeder
ranges between O, b to 1.0 kg/Mw(e). The estimated fuel doubling time.
for the MSER therefore falls in the range of 8 to 50 years. This is com-
parable to estimates of doubling times of 7 to 30 years given in fast -
breeder reactor design studies. .

From the point of view of long-term conservation of resources, low
specific inventory in itself confers an adventage upon the thermal breeder.
If the emount of nuclear power grows linearly, the doubling time and the
specific’ inventory enter symmetrically in determining the maximum emount
of raw material that must be mined in order to inventory the vhole nuclear
system. Thus, low specific inventory is en essential criterion: of merit
for a breeder, and the detailed comparisons in the next section show thet
a good thermal breeder with low specific inventory could, in spite of its
low breeding gain, meke better use of our nuclear resources than a good
fast breeder with high specific inventory end high breeding gain.

The molten seltuapproach to a breeder promiseS‘to satisfy the three
criteria of technical feasibility, very low pover cost, and good fuel
utilization. Its development as a uniquely promising’ competitor to the
fast breeder is, we believe, in the national interest. :

It is our purpose in the remainder of this report to outline the
current status of the technology, &snd to estimate what is required to
develop and demonstrate the technology for a full-scale thermal breeder
based on molten fluorides. - o

 

¥Total Kllograms of fissioneble material in the reactor, in storage
end.in fuel reprocessing and refabrication plents per megawatt of
electric generating capacity o , . L

—

O

§ N oo
‘i)( e

- Y .
& A

P

n

, "'EUEL-UT_ILIZATION COMPARISON
- Growth of Electric Generating Capacity

 The importance of good fuel utilization can be shown simply as
follows. A projection of the rate of growth of the.electrical generating
capacity in the U.S. is presented in Table 1. Numbers through the year
2000 were based on estimates developed by the Federal Power Commission

- and the AEC for the Report to the President in 1962 and were the nuclear |

capacities updated to reflect the present rapid growth of nuclear electric
capacity. The total capacities for the years beyond 2000 were based, in
Case A, on continued growth at the exponential rate of about 5% per year
and, in Case B; on continued growth at a linear rate of 100,000 Mw/yr--

the rate at year 2000. In Case B, the rate of expansion of total electrical

.generating capacity would be down to about 2% per year by the year 2030.

The nuclear capacities for the years beyond 2000 were extrapolated on the
basis that all new generating capacity after about 2020 would be nuclear.

Table 1. Electric Utility Generating Capacity

 

R

 

 

Total Capaoity Nuclear Capacity

Yoar | (1000 Mw) _( 1000 Mw) Percent

' Case A Case B Case A Case B Nuclear

1965 2ko 240 1 1 0.4
1970 330 330 n* nf 3
(1973)  (390) (3%) (36)° (36)2 (9)
1980 580 580 - 1ho® o® o
1990 1000 1000 390 390 39
- 2000 1700 1700 . 800 800 w7
2010 2900 2700 . 1700 1500 ~60
2020 5000 3700 3400 2500  ~70
2030 . 8600 4700 . 7000 - 3800 ~B0

 

- aProject.ions based on present rapid rate of sales of
g nuclear plants. Original nunbers were 6.8 for 1970 and
75 for 1980. Numbers for 1973 were not in the original

- projection but are based.on the present sales picture and
. lend support to the higher number for 1980. '

'Case A - exponential growth continued at rate of about 5%'.'
. per year'beyond 2000.

Case B - growth linear after 2000 at a rate ‘of 100,000 Mw
' per year. -

 
 

10

I
Nuclear Fuel Resources

Nuclear fuel resources estihated to be available in thé U.S5. to
support this expansion of the nuclear power industry are shown in relation
to cost in Teble 2. .If we define low-cost resources as those obteinable

- 0.

Table 2. U.S. Nuclear Fuel Resources (

 

-Cost Reasonable Assured ' Total Resources
($/1b Uz0s ‘Resources (thousand (thousand short
or ThO2) short tons of oxide) tons of oxide)

 

Uranium Resources

 

 

5 to 10 195 (L75%) 800%
10 to 30 Loo- 1000
30 to 50 5000 | 8000
50 to 100 6000 15,000
100 to 500 500,000 2,500,000 ]
Thorium Resources )
5 to 10 , 100 S 400 ¥
10 to 30 100 200
30 to 50 3000 10,000
50 to 100 8000 - 25,000
100 to 500 1,000,000 3,000,000

 

¥Includes all uranium delivered to AEC to date.

for less than $30 per pound, then our total low-cost resources are be=-
lieved to be 1.8 million short tons of UsOg, containing about 10,000
tons of recoverable 235U, and 600,000 short tons of ThOa. '

[

FuelJUtilizationnCharacteristics of Con%erter Reactors:

The efficiency of fuel utilization is.determined by the quantity of
Ua0g required to provide the total inventory of fissionable materiel
associated with the reactor per megawatt of electrical generating capacity
and the guantity of Ualg required per year per megawatt of electrical
generating capacity to provide for burnup of fissionsble materisl. These 2
requirements are listed in Table 3 for several types of reactors. The fiuj
reactors are more ‘advanced than are being built today, but the performance
d‘ (‘ . ‘ o | +) :’,50"",

" Table 37.‘3 "F.‘u‘el-Use Characteri‘stricé' of Several Types _of‘ Con‘ver_ter‘ Reactors

 

Specific Inventory o ' Annnal Consmnption :at 0.8 Total Load 'Fact‘.orb‘

 

 

- Reactor Type kg fissile /short tons U303) /short tons ThO \ kg fissile\ ‘short tons Usoa\ /short tons ThO-\

| Mw(_e) \ 1000 Mw(e)

"

 

\ 1000 Mu(e) N\ mele) /i 1000 Mw(e) /% 1000 Mw(e)

BRor AR 2.3  3506'7‘?@'- Y
I 1 2 260 ST 03h | L L
weR b 0 %0 o007 | 5 15
HIOCR-Th a;fl 520 - 130 o022 b8 o7
CEIGR - ‘- 3.1‘ _ 610 | 9% 011 o 0.8

VAeR 1.0 . 220 w0 0.5 o 1

 

Includes total inventory in reactor, fuel processing, fuel fabrication and storage.

bBased on recycle of plutonium.

Tt
 

12

indicated should be attainable within & few years, except possibly for
the hypothetical Very Advanced Converter Reactor, vhich has & much lower
specific inventory and & conversion ratio approaching one.. The latter is
included to show what greatly improved "advanced converters" or high-
performance near-breeders might accomplish. In the studies from which
the data were taken, the reactors were generally optimized to obtain the
lowest power cost from low-cost fuels. Recycle of plutonium is assumed
in estimating the burnup. Optimization for use of higher cost fuels
would have resulted in better, but not greatly better, fuel utillzation
and higher power costs.

Fuel Resource Requirements with- Converter Reactors'

The deta from Tebles 1 and 3 were used to obtain the curves in Fig. 1.
The assumption was made that only boiling or pressurized water reactors
would be built until 1976. Beginning in 1976 sdvanced converters associ-
ated with a given curve would begin to be built and by 1988 all new reac-
tors would be advanced converters. Each reactor built was assumed to have
8 life of 30 years. = '

The amount of uranium required for the inventory and the total burnup
to any given date is shown in Fig. 1 along with the total estimated re-
sources and the total cost of obtaining those resources. The fuel require-
ments for pressurized and boiling water reactors do not differ appreciably
and would require the mining of all our reserves costing less than $30 per
pound by shortly after the year 2000. If the industry continues to expand
as projected and the estimate of the availability and cost of the fuels
is reasonably accurate, all the fuel available for less than $50 per pound
would have to be mined by 2030 at a cost of about $700 billion. -The
advanced converters presently proposed will buy 5 to 10 years' time in
uranium reserves over the pressurized and boiling water reactors.

Further extension by converter reactors would require development of
a reactor--probably of a completely different type--with & much lowver
specific inventory and a higher conversion ratio. Even with such a very
advanced converter, the total domestic uranium resource, available for
less than $50 per pound UzOg, would be consumed by about 2050.

Figure 1 does not give the whole picture. A power reactor should
run dependably and,profitably for about 30 years, so when & reactor is
built, we, in & sense, commit a& fuel supply for 30 years. For the reac-
tors and growth rates used in meking the curves in Fig. 1, the total
commitment at any given time is about the same as the total shown for
_the inventory and burnup 10 years later. Reactors built as late as 1990
in an "all-water-reactor economy” would be fueled initially with uranium
costing as little as $10 per pound UsOg. However, the cost of fuel could
be expected to rise to $30 per pound of UaOg during the life of the plant
if there were no further expansion of the power industry, and to $50 per
pound if the industry continued to expand rapidly.

L.
[

b

)

i
-
.

: $30/1b Us0g Ǥ50 billion

$10/1b U30e.
~$10 billion

Resources Mined (thousand short tons)

8

—- —- _crase‘B

 

10 :

1970. . 1980 . 1990 2000 . 2010 2020 = 2030
- Fig. 1. Fuel -ReQuiréd, for Inventory and Current Burnup in Converter

Reactors. ' Co S - :

 
 

1k

The ThOz commitment is &bout the same for the HWOCR, HTGR, and the

VACR. The light water breeder reactor has a much greater thorium inventory.

In all cases the thorium inventory is several times the 30-year burnup, so

the amount of thorium required at any time is close to the total commitment.
Although much less thorium is required then urenium, the low-cost reserves

are smaller and would be used in inventory by 2010 to 2030.

The effect of the cost of UzOg and ThOz on the cost of power is shown

in Table L4 for the reactors and the corresponding inventory and consumption

numbers from Teble 3. These costs are only the costs associated with the
raw materials and do not reflect the higher enrichment, febrieation, pro-
cessing, and other costs that invariebly eccompany inereases’ in raw mate-
rial cost. They are, however, for reactors that have not been optimized
for use of ‘high-cost resources. All except the very best converter reac-
tors would suffer heavy penalties if the UzOg cost were to rise to $30
per pound, In the thorium reactors, the consumption is small, and for
those reactors with low inventory the use of high-cost resources has only
a small effect on the power cost. The light water breeder reactor would
incur a considerable cost penalty in an era of high-cost thorium.

Fuel Utilization Characteristics of Breeder Reactors

 

The effectiveness with which & breeder reactor can reduce the total
resource requirements depends on the specific inventory and doubling time
of fissile material in the breeder system, the growth rate of the nuclear
pover industry, and the capacity in converter reactors at the time the
breeders begin to be used for essentially all new capacity. Character-
istics taken from studles of oxide- and carbide-fueled fast breeders and
of a molten-salt-fueled thermal breeder are presented in Teble 5. The
estimated doubling times vary from T to 30 years for the fast breeders
and from 8 to 50 years for the thermal breeder.

Fuel Resource Requirements with Breeder Reactors

The total resource requirements¥* for a power industry in which only
water reactors are built until 1976 or 1986 and only breeders are built
after 1998 end 1998, respectively, are presented in Figs. 2 and 3. The
figures show the total resource requirements to depend heavily on the -
capacity in water reactors &t the time when breeder reactors are intro-
duced and, by comparison with Fig. 1, the great 1ncent1ve for expediting
the development of breeders.

The thermal breeder is clearly competitive with the fast breeders
in reducing the requirements for mined uranium. If the doubling time is
less than ebout 12 years, the maximm resource requiremeht depends more
on doubling time than specific inventory, so there is little difference

 

~ *Inventory in converter and breeder reactors, plus net consumption
by converters minus net production by breeders.

Q,

3]

i WM
) ( o Ay, o0

Table 4. Partial Effect of Us0g on Cost of Power'

 

P ( oy

 

~ ‘Contribution of Raw Material Cost to Power Cost (mills/kvhr)

 

 

 

 

 

 

Reactor Type — /I [ $30/10_ __Bo/w

' e  ;nventofy.—‘BurnuP7' Inventory = Burnup . Inventory  Burnup Inventory Burnup
G P ggos Requirements

BWR or MR 0.07 . 0.19 0.1  0.38 0.43 1.2 0.70 1.9

HWOCR-U = 0.0 0.0  0.07 0.21 0.22 0.66 0.37 1.0

IWBR 0.2  0.02 0.24 0.0k 0.67 0.1k 1.2 0.22

HWOCR-Th 0.0 0.07 0.1h 0.14 0.45 0.43 0.73 0.68

HIGR - 0.09 - 0.0k 0.19 0.07 0.58 0.21 0.94 0.3

VACR - . .  .0.03 0.02 0.06 0.03 0.19 0.10 0.31 0.16
| | ThOo Requirements

EWOCR-Th, HTGR, . - 0.0l 0.00 0.03 0.00 0.09 0.01 0.14 0.01

VACR e
INBR 0.05 0.00 0.11  0.00 0.33 0.01 0.53 0.01

 

aInventory chargedvat-lo% per year.

 

Gt
 

 

Teble 5. Fuel-Utilization Characteristics of Several Breeder Reactors
(Doubling time = 1/annual yield)

 

Specific Inventory

 

Doubling

 

 

fissile) (ahort tons Uaog) Breeding Time
Mw(e) 1000 Mw(e) Ratio (yr)
Liquid-metal-cooled fast breeder reactors ‘ _
Carbide fueled? 5 1100 1.b to 1.6 12 - 17
Carbide fueled?sf 2.h 520 1.h 8
Oxide fueled®»f h 870 1.2 to 1.3 18 - 28
Oxide fueledd:f | 3 650 1.2 to 1.k 10 - 20
Helium-cooled fast breeder reactor A S - o
Oxide fueled® h.3 ' . 930 | 1.5 | 12
. Carbide fueled | 3 650 1.6 T
Molten-salt thermel breeder reactor 0.4k to 1.5 87 to 320 1.03 to 1.08 8 - 50
MSBR with Pa removal | 0.7 150 1.07 14

 

®R. B. Steck (compiler), Liquid Metal Fast Breeder Reactor Design Study, WCAP-3251-1 Westinghouse
Electric Corporation (Janvary 1964).

Liqnid Metal Fast Breeder Reactor Design Study, CEND-200 Vbl.\I and II Combustion Engineering, -
Inc, (January 196L). .

“Large Fast Reactor Design Study, ACNP-6h503, Allis Chalmers (January 1964).

, . J. McNelly, Liquid Metal Fast Breeder Reactor Study, GEAP- hhlB Vbl. T and II, General
- Electric (January 19637.

A Study of a Gas-Cooled Fast Breeder Reactor3 Initial Study, Core Design Analysis and System
Development Program, Final Summary Report, GA=553(, General Atomic Division of General Dynamics
(August 15, 1964). |

fAn Evaluation of Four Designs of a 1000 Mie Ceramic Fueled Fast Breeder Reactor, C00-2T79,
Chicago Operations Office, U. 5. Atomic Energy Commission (December 1, 1964).

. C. | .. o 0,

1

 
 

 

44 ( ~ A) »

10,000 ; $50/1b UsOg 4700 billdon

 

$30/1b U505 ~450 btdlton

- Faat Breeders ‘ 'I.‘hérmailiBrééders —

. Resource Requirements (thousand short tons UsOs)

-—---- Case A ' --—---- - Case A

‘ .

  

1970 1990 2000 2010 2020 2030 1970 1990 2000 2010
. . , | ‘!ear - - \ . Year
Fig. 2. Total Fuel Requirements for Nuclear Power Industry Based on
Introduction of Breeder Reactors in 1976. -

 

  

2020 2030

 

yr

a} c =l

T
 

 

10,000, $50/1b Us0a  ~4700 billton,

    

 

50 yr
24 yr
50 yr
2h yr
1 yr

9 yr

430/ U550 Biitton

1000 O U0y~ eafl

100

. Thermal Breeders .

' Resource Requirements (thousand short tons UaOs)

Fast Breeders

Case A |
—--—-caaeB

————— Case A

| = e e Case B

10

 

1970 - 1990 - 2000 2010 2020 2030 ‘1970\ 1980 = 1990 2000 2010 2020 2030
' Year , | ' "Yeayr - B

- Fig. 3. Total Fuel Requirements for Nuclear Power Industry Based on
Introduction of Breeder Reactors in 1986. -
 

ay C >4

"»

ny

+)

119

between fast and thermal breeder systems., For longer doubling times, the
specific inventory assumes greater importance and the maximum requirements
for thermal breeder systems become increasingly less than those for fast
breeder systems with equal doubling times. Once the maximum requirement

- is satisfied, the fast breeders produce much larger amounts of excess

fissionable material.- Whether this is important depends on the need for
the excess material, ' | »

Figures 2 and 3 were based on starting the fast breeder reactors
with plutonium and the thermal breeders with 27U, The fast breeders
require an inventory of 3 to 5 kg of plutonium per megawatt of electric
generating capacity, and the PWR's and HWR's produce 0.2 to 0.3 kg of

plutonium per year per megawatt of electric generating capacity. The

growth rate of the nuclear generating capacity is T to 10% per year from
1980 to 2000. The converters and the breeders coming into operation
would be able to provide the inventory for high-performance .fast breeders
but would fall rapidly behind if the breeders were to have doubling times
longer than about 12 years. Additional thermal converters or fast con-
verters would have to be built or the breeders would have to be fueled
initially with 25U. This ‘could add significantly to the resource re-
quirement and the fuel cycle costs during the period of conversion to
operation on plutonium. T ‘ ,

Thermal breeders are also likely to be Pueled initislly with 235U
to produce an inventory of 233y, However, the conversion time is only
about one year and the additional resource requirement and the cost
penalty are small.

 

COST-OF~POWER COMPARISON
Capital Costs

Although moltenesalt thermelrbreeder reactors are competitive with
fast breeder reactors and superior to the converter reactors with respect

to the efficient use of nuclear fuel resources, they must also produce

power for a&s low or lower cost. No large molten-salt reactors or fast
breeders and few large advanced converters have been designed in detail,

8o most of the costs must be educated estimates based on comparisons of

the reactor systems and- judicious use of information from reactors that
are. being built. Such a comparison was made of several- advanced con-

- verter reactors and reported in ORNL-3686.2 The results are summarized
i”in Table 6, A comparable estimate of costs for a large molten-salt

thermal breeder reactor, made by the same people and reported in ORNL-

-'_3996 (ref. 3), is also included in the table, along with the fuel cycle

costs from several. studies of fast breeder reactors. Capital costs were
not estimated in the fast breeder studies. In all cases the costs in

| the table are for investor-owned utility plants which carry a 12% per

year charge on investment in. plant and 10% per year on inventory of fuel.

The comparisons show that the capital cost of a large power station
containing a melten-salt breeder reactor should not be much different

 
 

Table 6.

A Compariscn of Fetimated Cosats for Breeder and Advnnced Converter Reactorl _ -

Based on Investor-Owned Utilities Charges

 

 

 

 

 

Advanced Converter Reactors u;:lt.:n o Fast Breeder Reactors B
oo HWR - IWOCR Therml
LER » WCAP  CEND  ACNP OFAP . OA
R T FIOR 8GR U m T T hemcter 32511 200 6403 W8 9551
Cost for 1000-Mv(e) pover plent, $ million
Direct costs 9k wee 83 9% 88 96 86 &2 8
Indirect costs 59b o 35 39 37 %) . 34
Total capital 133° 153 118 132 12% 13 121 116 113
Speciel fluide o 0 0 o 27 33 1% 13 5 0 0 0 0 0
Fuel proceui.ng phnt o 0 o 0 0 0 0 o % 0 0 8 0 0
Pover costs _(mnll/kwh:) . . L L o
cdmm cost ’ 2.3 2.3 2.0 2.2 21 2.3 2.1 2.0 - 20 (2.3) (2.3) (2.3) (2.3) - (é.j}
Operating cost 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 e , o
Fuel cycle cost (t) . ‘ . _ | ‘
Fabrication 034 0.6l 0.26 020 031 0.22 0.% 0.5% =~ 031 026 016 034 0.3
Bumup and ].o.fle‘ 0.99 0.40 0.20 °o97 0.81 0.1 002’ 0059 0.01 : 0.02 0.02 mae - L me-
Processing B 0.20 0.27 0.19 0.19 0.25 0.2% 0.17 0.14 0.19° 0.13 0.12 0.5° 0.19 0.17
Bhipping O-O} 0-05 0.05 0.02 0003 0.0’I» O.CE o.oh one o OOOh 0-” - o : -
Inventory 0.2k 0.92 0.51 0.27 0.10 0.3 0.09 . 0.30 0,15 0.678 0.33% o058 o o,
Intere‘t on \'Ol‘king upim O-W Oow 001k o.m 0105 . 0-0’! 0-06 OOm : -ao o 01 0.11 0 O’G 0005 0.0’&
Subtotal 1.67 2.30 135 1.6 L5113k 101 L8 035 0 1.8 . 0.89  1.28  0.96 1.26
Pu or 2%y creait 0.20 0.2k 0 0.4 0.3 0 025 0 010 0.5 04l 032 019 . 0.%
Net cost L7° 21 Lk L5 L2 13 0.9 15 0.3 06 05 1.0 08 0.7
Special fluids inventory and 0 0 0 0 05 0.6 03 03 0.1 0 0 6 o 0
replacement ‘ . ' ‘ ' s _ ' T
Total power cost L3° T 37 - kO b1 k5 3.6 k1l 2.7 (2.9) (2.8) (3.3) (3.1)  (3.0)

 

».

a. Included because plant 1s uimiinr'to 80d1un-cooled fast breeder plants. .

power plants, the basis for the numbers is the same as for the other

converter reactors end for the MSER so they are used for this. conpanldn.
- The niumbers 4o not differ much from preliminary results. of recent studies

of normalized costs.’

Although these numbers l.re higher than present bid prieeo for large nuchnr

LG Includel capital charge on proceuing plmt.

4. Adjusted to 10% charge for investoromed utilities’ to bo
_consiatent with other studies.

e. ICapitaJ. costs‘takentobe thgmum

‘ f. Tuel cycle cost is 50-ye§r averaged cost. ruel eych cont tor"" C

aqumbrlun breeder cycle is 2.k nius

.02

 
 

'y C ¥

v

AY

9

21

from one containing a thermal converter reactor. We believe this is a
reasonable conclusion. . The molten-salt reactor uses high-nickel alloys--

~ vhich are. more. expensive than stainless steels--for structural material,
- uses expensive graphite in the core, has an intermediate heat transfer

system'between the reactor primary system and the steam system, and

_ requires special. provisions for remote maintenance of radioactive equip-

ment, However, the salts are good heat transfer fluids with high volu-

metric heat capacity, ere chemically stable at high temperature and, we

believe, at very high power density, have low vapor pressure, and can be
used with large temperature differences without mass transfer difficulties.
They do not undergo violent chemical reactions with air or water. The

‘primary and secondary systems can be compact and, except for parts of the

steam generators, can be built for low pressure. The reactor can be
fueled while at power by means of relatively simple equipment, and the
amount of excess reactivity can be kept small. The plant can operate

~at the highest thermal efficiency obtainable with modern steam plant

practice, so the cost in dollars per electrical kilowatt can be low even
though the plant nmay have more equipment and the dollars per thermal kilo-
wvatt may be higher than for a weter reactor.

Operating Costs

In Teble 6 the operating costs for the molten-salt reactors are

‘shown to be the same as for the converter reactors. Most of the operating

costs do not vary ‘much with type of reactor. We have not studied the
operation and maintenance enough to know whether an appreciable cost
penalty results from handling of the larger quantities of radioactive
wastes and from maintenance of the more-than-normally radiocactive equip-
ment in & molten-salt reactor plent, so none was included here. Several
million dollars was included in the capital cost for Spec1al maintenance
equipment. \

Fuei Cycle and Total Power Costs

- Tablé 6 shows that the fuel cycle cost for a molten-salt thermal
breeder reactor is lower’ than for any of the converter or fast breeder

'reactors. The molten fuel and blanket salts can be reprocessed continu-
ously or semicontinuously'by simple physical and chemical processes,
‘such as vacuum distillation and: fluoride volatility, in & small plant .

connected directly to the reactor. Fuel fabrication and shipping costs

~are eliminated; burnup cost (thorium) is negligible; the inventory
charges are minimal the credit for bred fuel is small. All these com-

. bine to ggoduce very'low fTuel cycle costs that depend very ‘little on the
~ sale of 3U. The contribution of the mined ThOg and Usoe costs to the

total power cost is small, so the increase in povwer cost in going from

| ' the present low-cost resources to $50-per-pound resources should be less
~than 0.3 mill/kwhr. The very low fuel cycle cost results in the molten-

salt resctor having an estimated power cost that is sdbstantially lower
than for any of the converter reactors.

 
 

 

22

If one accepts, in the ebsence of estimates, thet the costs for
building and operating large power plants containing fast breeder reactors
should not differ greatly from the costs for the other plants in Teble 6,
then differences in power costs depend'primarily on differences in fuel
cycle costs. According to the numbers in the table, the fuel cycle costs
and the total power costs for the fast breeder plante are mostly lower
than for the converter plants but higher than for the molten-salt thermal
v breeder plant. o |

How the molten-salt thermal breeder and the fast breeders compare
depends strongly on such characteristics of the fast breeders as the
relationship between the plutonium inventory, the breeding gain, the
charge assessed egainst the inventory, and the value of the excess plu-
tonium produced. These factors can be so adjusted that a fast breeder
with & very short doubling time could have negative fuel cycle costs.
In vievw of the many uncerteinties, we interpret the data in Table 5 to
indicate primarily that a molten-salt thermal breeder plant could produce
pover at & cost competitive with the cost of power from a fast breeder
plant and with far less dependence on the sale of fissionable materisl.
The molten-salt thermal breeder is clearly a strong competitor to the
fast breeder for achieving the goal of producing pover at low: cost with
good fuel utilization.

1000-Mw(e) MOLTEN-SALT THERMAL EREEDER POWER FLANT

Studies of the conceptusal design of & lOOOéMw(e) molten-salt thermal
breeder power plent (MSBR) end of some alternatives or improvements are
reported in ORNL-3996, ORNL-4037, and ORNL-4119. Results of the studies
are summarized here end in some instences are adJusted to inco:porate '

- more recent information.

The MSER reference design is & two-region, two-fluld system with
fuel selt separated from the blanket salt by graphite tubes. The fuel
salt consists of uranium fluoride dissolved in a mixture of lithium and
beryllium fluorides, and the blanket salt ‘is & thorium fluoride — lithium
fluoride mixture of eutectic composition. The heat generated in those
fluids is trensferred in a primary selt-circulating system to & coolent
salt in a secondary circuit which couples the reactor to a supercritical
steam cycle. Fuel and blenket are processed on site by'means of fluoride
volatility end vacuum distilletion processes. _

A design called MSBR(Pa)-is e fevored veriation of the MSER. It is
the same as the reference design except that the blanket salt is processed
- to remove protactinium on ebout & half-day cycle. This results in improved

- performance through & higher breeding ratio, & smaller inventory of . .
fissile material in the blenket, and a considerable reduction in the in-
ventory of blanket salt. , ‘ _
&y

H

23

Two methods of removing protactinium from fiuoride melts have been
tested on small scale in the laboratory. In one, PaOz was shown to pre-
cipitate on ThOz that had been added &s & solid to a molten fluoride
salt, In the second, protactinium was extracted from a fluoride melt

by molten bismuth with thorium metal as & reducing agent. The chemistry
of these processes is favorable, so further work should provide an in-
- organic ion exchange process or & liquid-metal extraction process for
.removing protactinium continuously and inexpensively from the blanket

salt of a breeder reactor.

Because the designs are so similar the MSBR and MSBR(Pa) are treated
below as one plent. Characteristics for both are reported where they
differ.

| Reference Plant Design

Fuel, Blanket, and Coolant Salts

Fuel salt for the reactor is a ternary mixture oonsisting of about
0. 3%* UFy, 65.7% "LiF, and 34% BeFp. This salt is similar to the fuel
in the Molten-Salt Reactor Experiment. A salt containing 27% ThFg, T1%
LiF, and 2% BeFo is proposéd for the blanket. A mixture of 48% NaF, W%
KF, and 48 BF3 is the favored coolent salt because of its low 1iquidus
temperature and low cost. Estimates of the physical properties of the
salts -are reported in Table 7

Flowsheet |

A flowsheet for the lOOOéMw(e) plant is presented in Fig. L., Fuel

-1s pumped through the reactor at a rate of about Lk,000 gpm, entering

the core at 1000°F and leaving at 1300°F. The primary fuel system has
four loops, each loop having a heat exchanger and & pump of 11,000-gpm
capacity. The blanket system has four pumps and heat exchangers, smaller
but similar to the components in the fuel system. Blanket salt circu-
lates through each of the four- loops at a rate of 2000 gpm, entering the
reactor vessel at 1150 F and leaving at 1250 F. :

 Four 1k, Ooo-gpm pumps oirculate the sodium fluordborate coolant

. salt through‘the shell sides.of the primary heat exchangers. The salt
~ enters at 850°F, leaves at 1112°F, and then passes through the shell
- sides of the blanket heat exchangers where it is further heated to

1125°F, The coolant then passes in parallel through sixteen once-through

' boiler-superheaters and eight steam reheaters._

 

*¥A11 values are in mole %.

 

 

 
 

 

Table 7. Estimated Physical Propertiee of MSBR Fuel,

Blanket, and Coolant Salts®

 

"Blanket Salt

 

Fuel Salt Coolant Salt
Composition, molo ¢  65.7 LiF-34.0 BeFz~ 7L LiF=2 BeFz— 48 NaF-k KE—
S . ‘7 0.3 UFe | 2T ThF4 48 BFa
Liquid temperature,_ °F o 852 ' 10%0. w700
Reference temperature for properties, °F - 1150 5_1200 988
Density, 1b/ft® | B 127 217 125
Viscosity, 1b/ft-hr = R 19 38 12
Thermal conductivity, Btu/ft-hr- F . 0.6 0.4 0.5
0.22. "0,37

“Heat capecity, Btu/1b-°F ] . - 0.55

 

83, Cantor, R. E. Thoma, J. W. Cooke, and H. W, Hoffman, Estimated Physical Properties of
MSER Fuel, Blanket and Coolant Salts, ORNL-CF-67~3-18 (Mareh 10, 1967). |

He

 
 
 
 
 
  
  

 

 

 

 

 

  

- FUEL SALT HEAT

 

Y JusE

)

§50p- 1000°F

 

 

   
  

 

 

 

 

 
 
  
   

20 f1 Veae

 

  

3600p- 1000°F |

!

|

1

1

1
)

   

1 14240

 

.| 850°F

—— v ——— -

RENEAT

|~ PREMEATERS

o
6920 |

 

 

 

 

 

.- FUEL
| BLAMKEY .
~ COOLANT. —=--m
 STEAM -
CHO

 

 

o0t

bl BtR/Ih
ommemnntF

D= ___Freers Volvi

Sl e 3
STl v
TS " . possh 4 ]
% ’} ‘ ¢ ;5 : ‘ .570.'550.F§—’
TBLANKET SALT - (T FUEL SALT \
DRAIN TAKKS / - \_DRAIN TANKS /

 

]
RS

T — .

1
1307.8%

 

sTeaM |

  

-

= i

1

I424l|_~35|5rI000‘ i
‘ 1

  
 

TURBINE

   

I‘i,ii
R
b b

 
 
    
      

- &) C b

1518.5h+ 540p-1000°

 

- GEN. '
4 527.2 Mwe
- Gross A

 

 

 

 

. GEN. |
S07.7 Mwe |
Gross ..

 

 

 

|
1
[
b
¥

 

 

SYSTEMS

CONDENSER B FEEDWATER|

 

 

35009-866°F

 

34759-695°F Y
16644

- BOOSTER

. pumps
COOLANT SALT\ = -
~ \_DRAIN_TANKS |

 

258 | g3

    
  

 

 

]
MIXING TEE

_ - PERFORMANCE
NET OUTPUT

GROSS GENERATION
BF BOOSTER PUMPS

- STATION AUXILIARIES

NET EFFICIENCY

REACTOR HEAT INPUT
NET HEAT: RATE

Fig. 4. Molten-Salt Breeder Reactor Flow Disgram.

3500p- 5509°

1000 Mwe
10349 Wwe
9.2 Mwe
25.7 Hws
2225 Hwt
7601 Bis/hwh
M9%

qc
 

 

 

 

':26 o : <

The steam.system is essentially that of the new Bull Run plant of the
TVA, modified to increase the rating to 1000 Mw(e) end to preheat the work-
‘ing fluid to TOO°F before it enters the boiler-superheaters. Use of the
supercritical steam cycle eppears to ease some problems of design of steam
- generators for molten-salt reactors and results in & tlsrmsl efficiency of
- gbout 45%. | _

Reactor Design

The MSER cell arrangement 1s. shown in plan in Fig. 5. On two sides
of the reactor cell are four shielded cells containing the boiler-super-
heaters end the reheaters; those cells can be isolated .individually for
maintenance. A cell for handling the gaseous fiseion products from the
reactor end two cells for processing the fuel and blanket sslts are ad-
Jecent to the reactor cell. Cells are also provided for decontamination
and storage and repair of radioactive equipment.,

An elevation of the plant in Fig. 6 shOWS‘the arrangement of equipment
in the reactor and coolsnt cells, and a more detailed view of the reactor
primary equipment is shown in Fig. 7. The reactor vessel is ebout 14 ft /
in diameter by about 19 £t high, is designed for 1200°F and 150 psi end
has & metal-wall thicknesses in the range of 1 to 3 1n. :

The reactor vessel contains a 10-ft-diem by 12-1/2-ft-high core
assenbly composed of 534 graphite fuel cells of a type similar to that
- shown in Fig. 8., Fuel from the entrance plenum in the reactor vessel
flows upward through the annulus and downward through the large central
passage in the graphite tubes to the outlet plenum. Fuel is circulated
from the outlet plenum.through the pumps to the heat exchangers and then
back to the reactor. 1-1/2-ft-th1ck‘blanket and & 3=-in.-thick graphite
reflector surround the core. The thorium salt circulates through the
‘blanket region, through the passages between fuel cells in the core, and
through the heat removal system outside the reactor vessel,

Values chosen for some of the MSER and MSBR(Pa) design parameters
are listed in Table 8, ‘

The reactor vessel and all other equipment that holds salt is madé

~ of Hastelloy N, a nickel-base alloy containing ebout 1T% molybdenum, T%
chromium, and fl% iron. This materiel is highly resistent to corrosion

by fluoride salts end has good strength at high temperature. The high-

_ temperature creep properties of Hastelloy N presently obtainsble commerci-
ally deteriorate under irradietion, dbut small changes in the alloy offer

promise of eliminating this deficiency.

The graphite 1s a high-density grade processed to achieve small pore
openings for low permesbility to szlt. Superior resistance to damage by
irradietion is important, but the core is designed to keep the flux gradi-
ents emall seross individusl pieces and to permit the grsphite to expand
or contract with little restraint.
) )

 

     

 

   
   

REMEAT STEAM '
O HP STEAM ?‘_. ) WASTE GAS FEEDWATER
1 H.P. STEAM

CELL

   
  
 

. FEEDWATER
o LP. STEAM

 
   

FUEL -
HEAT EXCH.

      
   

 

. COOLANT SALT.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

L2

 

 

 
      

L—l6 SUPERHEATERS |

BLANKET

_oscowmmwmm\; o
‘ -HEAT EXCH.

AND STORAGE

 

 

 

 

 

 

 

. CONTROL AREA—~

 

 

 

 

Fig. 5. Molten-Sal'b Breeder Reactor — Cell Arrangement, Plan Viefi.
COOLANT SALT
| PUMPS

- SUPERHEATERS—._

 

  

FUEL HEAT EXCH.

- CONTROL ROD DRIVE

BLANKET CIRCULATING
PUMP

 
  
     
 

FUEL CIRCULATING 7

 

L pLANKET REHEATERS
'HEAT EXCH

Fig. 6. Molten-Sal'b Breeder Reactor — Cell Arra.ngement Eleva.tion.

   

__GROUND LEVEL -

gz

 
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)

)

v

)

29

   

 

FUEL PUMP - BLANKET PUMP
YMOTOR , -\ MOTOR |

BLANKET
PUMP

BLANKET
HEAT EXCH

- FUEL
PUMP

10 FT. DIA.
CORE

REACTOR
VESSEL

SALT
DISTRIBUTION
PLEMIMS

.
.
S .
ey
frtil
AT ..

CELL HEATERS =

( | Fig. '7 V'Réac_:fbr_l?rimary Equipae’nt.
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    
  
  

 

 

 

 

   
  
  

 

 

 

 

 

 

 

 

 

30
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l { 6- } %1} siors
AR GRAPHITE TO
) METAL ERAZE
r-e RIS %
’-0° o
a0 13
'é OD XIA Lo
- | 1§ 0.0.x1} o,
TAPERED THREAD
-4 i
i

FUEL INLET PLENUM — é
f

N\

-

 

v

——FUEL OUTLET PLENUM
Fig. 8. Molten-Salt Breeder Reactor Core Cell.
¥ C o}

)

41

3)

31

Table 8,1JRéactor Design Values®

 

 

MSBR(Pa) MSBR
Power, Mw |
Thermal S 2225
Electrical 1000
Thermal efficiency, fraction 0.45
Plant load factor - 0.80
~ Reactor vessel
Outside diameter, ft 1
Overall height, ft ' _ : ~]19
Wall thickness, in. : _ 1.5
Head thickness, in. 2.25
Core |
Height of active core, ft 12.5
Diameter, ft. | | - 10
Number of graphite_fuel passage tubes - . 534
Volume, ft - | ' 982
Volume fractions | D _
Fuel salt | | 0.169 0.169
Blanket salt | 0.073 . 0.07h
Graphite moderator — 0.758 - | 0.757
Atom ratios . _
Thorium to ursnium = . 2 4o
Carbon to uranium - 5800 sihlho
Neutron fluxé core averege, |
neutrons/cm®+sec )
Thermal . - T.2x10%4 6.7 x 10
Fast 12,1 x 104 12.1 x 10%*
Fest, over 100 kev - | ‘3.1 x 10%4 3.1 x 104
 Power density, core average, ' :
kw/liter . o -
Gross e - 80
“In fuel salt e - , - b3
Blanket | | o B
- Radisl thickness, £t _ 1.5 .
Axial thickness, £t . | 2.0
 Volume, £t " B . 120
_ Vblume fraction, blanket salt 1.0
Reflector thickness, in. ,: 3
Fuel salt | L ol T o
- Inlet temperature, °f .. . . 1000
Outlet temperature, °F S B - 1300
Flow rate, ft/sec (total) 95.7

gpm o ; . k2,950

Continued

 

 
 

32

Teble 8 (continued)

 

 

MSBR(Pa)H MSER
Fuel salt (continued)
Nominal volume holdup, ft3
Core 166
Blanket 26
Plena 14T
Heat exchangers and piping Lok
Processing plant 33
Total . TT76
Nominal salt composition, mole %
LiF ' 65.7
BeFa 3k.0
UF¢ (fissile) 0.22
Blanket salt | -
Inlet temperature, °F 1150
Outlet temperature, °F 1250
Flow rate, £t3/sec (totel) 17.3
gpm TTE
Volume holdup, f£t°
Core T2
Blanket 1121
Heat exchanger and piping 100
Processing 24
Storage for protactinium decey 2066
Total 1317 3383
Salt composition, mole %
LiF T1.0
‘BeFp 2,0
ThF¢ 27.0
UFe (fissile) - 10,0005
System fissile inventory, kg T2k 812
System fertile inventory, kg 101,000 260,000
Processing data | ‘
Fuel stream
Cycle time, days k2 7
' Rate, £t3/day 16.3 1k, 5
Processing cost, $/ft3 190 - 203
Blenket stream ,
Equivalent cycle time, days
~ Urenium-removal process 55 23
Protactinium-removal process 0.55 |
Equivalent rate, £t per dey B
Uranium-removal process 23.5 DU
Protactinium-removal process 2350
Equivalent processing cost (based 65? 7.3

on uranium removal), $/ft>

 

Cdntinued
F

n

x)

¥

-

33

i,TenleAS (continued)

 

 

MSER(Pa) ~ MSER

Fuel yield, %/yr B N 7.5 ‘- 4.5
Net breeding ratio , T 1.07 1.05
Fissile losses in processing, ‘atoms 0.0051 " 0.0057

per fissile sbsorption- | A
Specific inventory, kg of fissile - 0.7T24 o 0.812

material per megawatt of '

electricity produced
Specific pover, Mw(th)/kg of D -

fissile materisl , . 3.1 o 2.7
Fraction of fissions in fuel stream 0.996 0.987
Fraction of fissions in thermal-nentron | -

group o . 0.815 - 0.806
Net neutron- production per fissile .

absorption (xE) | v - 2.227 2,221

 

%Mmis table and others teken fromHORNL-3996 have been revised to
1nc1ude the effects on inventories of a reduced thermal conductivity of
the fuel. salt. : , :

Heat Exchange>§ystems

The fuel heat exchangers are of the tube and shell design and are
combined with the pumps as shown in Fig. 9. Fuel salt from the reactor
flows into the impeller of the pump and is discharged down through the
tubes of the inner bundle. It then flows upward through the tubes of

the outer bundle and back to the reactor core. The coolant salt enters

_the shell at the bottom, flows upward along the outer wall, then through
the tube bundles countercurrent to the flow of the fuel salt and out
through the center pipe. . *;va — o -

The blanket heat exchangers transfer only a small fraction of the

~_heat, but they pass the full flow of coolant from the fuel heat exchangers.
-+ They are similar to the fuel heat exchangers end are designed for single-

pass flow of coolent on ‘the shell side, although twe-pass flow is retained

| -for'blanket salt in the tubes.;_-_=

Fuel and blanket pumps are sump-type pumps bnilt into the upper heads

'~,‘of'fhe heat exchangere. ‘While this .complicates the design of some of the

equipment it reduces the salt inventory (particularly in the fuel system),
the amount of piping, end some of the stress problems during heating and
cooling of the systems. Concentric piping is used between the reactor

 

 

 
 

 

'UPPER BEARING— =~~~

SEAL
SHIELDI

6-0°'0n TANK

M.S.BEARIN
IMPELLER —

FUEL TO
REACTOR

OUTER TUBES
$794 €% 0D

-t

INNER TUBRES
4378 @ % 0D

sn.eron

 

FUEL DRAIN

 

-—

Fig. 9. Molten-Salt

    
   
 
 
  
   
 
  
  
 
  

"GAS CONN.

START-UP LEVEL

IGH OPR. LEVEL

OW OPR. LEVEL

S .
ILL ¢ DUMP

Breeder ReaC'bdr lPr‘imary Heat Exchanger and Pump.
a¥

n

X}

»

35

- vessel and the heat exchangers for the same reason. The fuel heat ex-

changers and pumps are below the core so the fuel salt in the core will
drain quickly into tanks, where it can be cooled more easily, if the

punps stop.

The boiler-superheaters are. long, slender, U-tube -U-shell exchangers.

Coolent salt flows through the shell, entering at 1125°F and leaving at
850°F. Water preheated to about. 700°F enters the tubes at 3800 psi and

leaves as supercritical steam at 1000°F and 3600 psl.

Steam is extracted from the high-pressure turbine at about 550°F and
reheated to 1000°F and 540 psi before use in the intermediate pressure
turbine. This is accomplished by heating partly with prime steam in pre-

_.heaters and partly with coolant salt in reheaters.

, Since the freezing temperature of the coolant salt is about T00 °F,

. it seems desirable to preheat the working f£luid to almost TOO°F before

it enters the boiler-superheaters or reheaters. This is the purpose of

the steam preheaters shead of the reheaters. The prime steam from those

preheaters is injected into the feedwater in a mixing tee to heat the .
water to the desired temperature before it enters the boiler-superheaters.

Use of the supercritical steam cycle makes possible this matching of

: salt and feedwater temperatures. It is believed to reduce the thermal

cycling (and fatigue) of the tubes that would occur in the boiling regions

‘of the steam generators at lower pressure. The net thermal efficiency -

of the plant is about hs% and would be higher if higher temperatures
could be used effectively in the steam system.

Fuel and Blanket Processing

The primary objectives of the processing are to separate fission
products in low concentration from the other constituents of the fuel
salt and to separate bred fissile materiel in low concentration from the
other constituents of the blanket salt whlle keeping the losses and the
costs low. With the fluoride fuel and blanket.salts of the MSER, these.

. objectives cean be fulfilled by & combination of fluoride volatility,
vacuum distillation and. protactinium extraction processes. The process-

~ing is done continuously or semicontinuously in cell space adjacent to -

- the reactor; services and some other equipment required for the reactor

-~ are shared by the processing plant. _Shipping, long storage at the reactor

~ and reprocessing. sites, end refabrication of fuel and blanket are elimi-

. nated.. All these factors lead to. reduced inventories, im@roved fuel
.utilization, and reduced costs.;~ : , , _ . ,

The fuel salt for the MSER &nd the MSBR(Pa) is processed by fluoride
volatility to ‘remove the uranium and by vacuum distillation to separate

- the carrier salts from the fission products. For the MSER the blanket is

processed by fluoride volatility alone, The cycle time i1s short enough’

to maintain the concentration of fissile material very low. The inventory

of blanket salt is made large to keep the Pa losses small. For the MSBR(Pa)
the blanket stream is treated by a liquid-metal extraction process

 

 
 

 

 

36

or an exchange process to remove Pa and 2337 on & very short cycle, In
this case the fissile inventory in the blanket and the blanket salt :
inventory can be kept to a minimum , .

- Principal steps in the processes are shown in Fig. 10. Small streams

" of core end blanket flulds are withdrewn continuously from the reactor .
and circulated through the processing system. After processing, the decon-
teminated fluids are returned to the reactor at convenient points such as
the storage tenks. Inventories in the processing plant are estimated to .
be about 5% of the reactor fuel system inventory end less than 1% ‘of the
blanket inventory. _

The fuel snd'blanket processing plants are intended to operate cone
tinuously in conjunction with the reactor. However, the reactor can con-
tinue to operate when &1l .or part of the processing plant is shut down for -
maintenance., During & 30-day’ interruption in processing of the blanket,
the increase in concentration of 2337 4in the blanket salt would produce an .
increase of less than 20% in the emount of heat generated in the blanket.
Since 23U would not be availeble from the blenket, the burnup in the core
would have to be compensated by supplying fissile msterial from & reserve.

. Interruption of the processing of the fuel stream.would cause the
fission product concentration in the fuel to increase. Fissile materisl
would have to be edded to compensate for burnup and for the:gradual increase
in poison level. During periods of operation without processing, there
would also be & gradual decrease in the breeding gain. The decrease would
be less than 0.02 in 30 days. ' - '

Capital-Cost Estimates
Reactor Power Plant

Preliminary estimates of the capital cost of a 1000-Mw(e) MSER power
station indicate a direct construction cost of about $81 million. After
applying the indirect cost factors used in the advanced converter evalu-
ation,® the estimated total plent cost is $115 million for privete financ-
ing end $111 million for public financing. - A sumsary of plant costs is
given in Teble 9. The conceptual design was not sufficiently detailed to
permit a completely relisble estimate; however, the design and estimates
were studied thoroughly enough to make meaningful compariséns with previ-
ous converter-reactor-plant cost studies. The relatively low capital
cost results from the small physicel size of the MSBER and the simple con-
trol requirements. The results of the study encourage the belief -that
the cost of an MSBER power station will be as low as for stations utilizing
other reactor concepts.

The - operating and maintenance costs of the-MSBR were not estimated.
Based on the ground rules used in Ref, h these costs would be 0. 3h
mill/kwhr(e) |
) ( ) %) n

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   
  

 

 

 

 

 

 

 

  

 

 

 

 

  

 

 

 

 

 

  

 

 

 

 

 

 

  

 

 

 

 

o L UFs RECYCLE TO REACTOR . ‘
R | STV
VY /;0}// ///4/";“% : ' 7 “ ' X777, 7777
| VW 777, o ///soaagi/// 4
o0, %%/,W& //'W% . N A
' NN A / Z S S
Lt -1~ 7 NN ]
e ‘ ‘ /" Nof // ; %AQFZJ; ; ~70*C
| e | el dpine |
b | | 777
- /msre- 4 j i /,sxc_sss 7
. [7STORAGE 73 - ' e e : . FPRODUCTION
e MR S " waste 77 (L LLL
. UFgt . . KLLelis _ : o ‘ "STORAGE : '
akEup | VORATLEFP o o _ [ NaF/MgF,/FP
UF/BeRyTHF, o f T | UFg + - |
L T o VOLATILE FP MAKE UP’
: : ‘ LiF/BeF,
I | _ . ‘ o | _ o \ 2 Yy ' 2
7T -E’”?l?”””//’/ 7////7777/'//7//
. F/Eanl-y% 7 FLUORIDE % ‘ . roR ,/co;u:dgg%és4 ; %/”fi//r D'/S.T'LL‘“E/ |7 U_FG"UF‘/; / /’f
7 MAKEVD N/ 5‘!05%“}'"2 MSSR sPENT [ Fp pecay A voLaniLity 2} /| DisTiLLaTion | - LiFBeF, /|2 ) repucTion [ ;urmnon -
r . 7V /A L : FUEL -~ "V s s 2/ | /) Assose 7./ | ~1000°¢ |/ | /~s00%C 77, [/ |7, 550-600°C / z
// %///55/00/(;///% ‘ ///// / | @‘5 DAYS é{‘:wz"fs/j !mmfl/;//////; /////////
0722207 | A N0
. d L R i A _ : ““ | '
LaFIBcFZIThFdFP R L L ETR;SAFRPF T:;DUCEU METALS
S /N __MSBR(Pa) b Cr. Fo Ni
-V s S P 4 L.
St 7o) L on nsrt
7 pemon ] s s grorsce 7 |
FERTI ‘ S LLLLLLL | - LiF /BeF, /UF, RECYCLE
LE STREAM RECYCLE A __ - ‘ —
_ FERTRLE STHEAM RECYCLE SALT DISCARD FOR

FP REMOVAL

Fig. 10. MSER Core and Blanket Processing Schene.

LE
 

 

38

Table 9. Preliminary Cost-Estimate Summary® for a 1000-Mw(e) Molten-Se.lt

Breeder Reactor Power Station [MSER(Pa) or LfiBR]

 

 

ifidustry.
b

Containment cost is included in Account 221.3.

®See Table 3 for these costs,

8Estimates are based o 1966 costs for an established molten-salt nuclear power pla.nt '

(in thousands of -
Commission dollars) i
Account
20 Land and Land Rights 360
21 Structures and Improvements
211 Ground 1mpr6vements 866
212 PBuilding and stmctures
.1 Reactor building 4,181
+2 Turbine bullding, auxiliary bu.tldj.ng, and feedwater 2 ,832
heater space . .
.3 Offices, shops, and laboratories 1,160
4 Waste disposal huilding 150
.5 Btack 76
.6 Warehouse ko
T Miscellanecus _ 30
Subtotal Account 212 88
‘Total Account 21 9,335
22 Reactor Plant Equipment
221 Reactor equipment
.1 Reactor vessel and mtema.ls 1,610
.2 Control rods 250
«3 Shielding and containment 2,113
.4 Heating-cooling systems and vapor-suppression 1,200
system -
+5 Moderator and reflector 1,089
.6 Reactor plant crane 265
Subtotal Account 221 6,527
222 Heat tr:a.nsrer eyatgms
.1 Remctor coolant system 6,732
.2 Intermediate cooling system 1,947
+3 Steam generator apd reheaters 9,855
.4 Coolant supply and treatment 300
Subtotal Account 222 ) - 18,832
223 Muclear fuel handling and storage (drsin tanks) 1,700
224 Ruclear fuel process:lng and tabrication {included :I.n (c)
' mel-cycle costs) .
225 Radioactive waste treatment and dispose.l (off-gas 450
system)
226 Instrumentation and controls k,500
227 Feéfiwater supply and treatment 4,051 .
228 BSteam, condensate, and feedwater piping 4,069 -
229 Other reactor plant equipment (remote mainte.fia.nce) zzoood
Total Account 22 15,129

WA ( .\)u.l
"

39

‘Table 9. (continued)

 

 

F;gf'::'l (in f.hgf:::fms of
Commission - dollars)
Account
23 Turbine-Generator Units
| 231 Turbine-genératér wnits 19,174
232 Circulating-water system 1,243
23__’5A Condensers and auxiliaries 1,690
23k Céntfal lube-oil system | 8o
235 Turbine plant instrumentation 25
236 Turbine plant piping ' 220
237 Auxiliary equipment for genere.for 656
238 Other turbine plant equipment ___ee
; Total Account 23 22,523
24 Accessory Electrical
’ 241 Switchgear, main and station service 500
242 Switchboards 128
243 BStation service transformers 169
24k Auxiliary generator 50
glfi DiBtributet_i 1tjems _2,000
. Total Account 2k ! 2,897
25 77 Miscellaneous | ___800
Total Direct Construction Cqst 80,684
Privs.te Financing
Total indirect cost 33,728
Total plant cost . . C1b, k2
Publ:lc Fina.nc:i.ng '
. © Totel indirect cost L 30,011
' Total plant cost® = - 110,695

 

q£Joes not include Account 20, Land. Costs. Land. is treated as a nondeprecia.ting capital item.

However, land costs were mcluded when computing indirect costs.

 
 

 

 

Fuel.Recycle Plant

The capital costs of the fuel recycle plant for processing 15 £t3/day
of fuel salt end 105 £t3/dey of blanket .salt in a 1000-Mw(e) MSER power
station were obteined by itemizing and costing the major process equipment
and by estimating the costs of site, buildings, instrumentation, waste -
disposal, and building services associated with fuel recycle. Teable 10
summarizes the direct construction costs, the indirect costs, and totel
" costs’ of the plant. The totel is $5.3 million. The opereting and main-
~ tenance costs for the plant include lebor, lebor overhead, chemicals,
utilities, and maintenance moterials. The totel annual cost is estimated
to be about $721;000, .which is equivalent to ebout 0.1 mill/kwhr(e)

A breakdown of these charges' 1s given in Table ll.r

Teble 10. Summary of Processing-Plant Capital Costs
o for a lOOO-Mw(e) MSER'

- ‘ _ W

 

 

Installed process_equipment T g 853,760
Structures and ifibrovements_ o 556, TT0
Waste storage 387,970
Process piping - 155,800
Process instrumentation | 272,100
Electrical suxilieries D 04,300
Sampling connections - | 20,000
Service and ut1lity piping | 128, 060
Insulation . | , 50,510
Radiation monitoring - "~ 100,000
 Total direct cost | $2,609,270
Construction overhead 782,780v
(30%- of direct costs) - | e—
Subtotal construction cost $3,392,050
Engineering snd inspection | o 848,010
-(25% of subtotal construction cost) ‘ '
Subtotal plant cost - $4,2L40,060
Contingency (25% of subtotal . 1,060,020
plant cost) 7

Total capitel cost | o $5, 300,000
¥

»

Table 11.

b1 °

~Summary of Annual Operating

and Maintenance Costs for Fuel
Recycle in a 1000-Mw(e) MSBER

 

Direct labor $222, 000
Lebor overhead’ - 177,600
~ Chemicals Al&,6h0
Waste'containers 28,270
Utilities 80,300
‘Maintenance materials | .
site 2,500
Services and utilities 35,880
Process equipment . 160,000
Total annual charges $721,230

 

 The capital and operating cests‘for'this nlant fiere the basis for

“deriving the costs of plents with other capacities. -

The relationship

of cost to volume of salt processed was estimated separately for fuel

and blanket streams to give’the:curves shown in Fig. 11.
 those curves were used in the fuel-cycle-cost optimization studies to
represent the effects of varying the plent size and, throughput.

Data from

~

For the MSBB(Pe) plent the processing methods and costs were the

same as those for the MSBR plant except for the blenket processing.

The

cost of protactinium removalifrom therblenket stream was estimated to be

C(Pe)

045'

l 65

of the total costs of fuel recycle in the MSBR(Pa) were based on the

curves in Fig. 11 for the blsnket stream, -

, Nuclear Performance and Fuel Cycle Analyses Ve

(1)

where C(Pa) is the capital cost of the protactinium removel equipment,
- in millions of dollars, end R 1s the processing rate for protactinium
- removal in thousends of cubic feet of blanket salt per dey. Calculations

 curves in Fig. 11 for the fuel stream-and on Eq. (1) combined with the

‘The fuel cycle cost and the fuel yield are closely related yet inde-
pendent in the sense that two nuclear designs can have similar costs but

significantly different yields.

The objective of the nuclear design

calculations was primarily to find: the conditions that gave the lowest
 

 

42

O,

| | - 'ORNL-DWG 66-7455
BLANKET STREAM PROCESSING RATE (ft*/day) o

2 ' 5 ' 10° 2. 5 los

 

 

AL COST OF CORE PROCESSING

/

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\,\\ I | N I N
2 S ———— f 2 T
'.g N N L -0PERATING COST OF CORE PROCESSING| | | | &
§ e | .\4 TN \ 1 1 : o 2
g VO IO ~C N T s .
N Ty 2
_ _ b=
| 1 \=£-CAPITAL cosToF | ¥
§ OPERATING COST OF —— ELANKET PROCESSING T| -~ 2
* BLANKET PROCESSING—"" \K | | 3
: — NG
2 . s . 2 . 8. 100

CORE STREAM PROCESSING RATE (ft¥day)

- Fig. 11. MSER Fuel-Recycle Costs as a Functlon of Processing Rates.
Fluoride volatility plus vacuum distillation processing for core; fluoride
volatility processing for blanket; 0.8 plant factor; 12%/yr capital charges
for investor-owned processing plant.
 

" C ”

»

»y

k3

'_ fuel cycle cost, and then, without appreciebly increasing this cost, the

highest fuel yield.

| Analysis Procedures and Basic Assumptions

The nuclear caleulations were performed with a multigroup, diffusion,
equilibrium reactor program, which calculated the nuclear performance,
the equilibrium concentrations of the various nuclides, including the
fission products,:and the fuel-cycle cost for a given set of conditions.

. The 12-group neutron cross sections were obtained from neutron spectrum

calculations, with the core heterogeneity taken into consideration in

the thermal-neutron-spectrum computations. The nuclear designs were
optimized by parameter studies, with most emphasis on minimum.fuel-cycle
cost and with lesser weight given to maximizing the annual fuel yield.
Typical parameters varied were the reactor dimensions, blanket thickness,
fractions of fuel and fertile salts in the core, and the fuel- and fertile-
stream processing rates.

The basic economic assumptions employed in obtsining the fuel-cycle

“costs are given in Teble 12. The processing costs are based on those

given in the previous section end are included in the fuel-cycle costs.
A fissile material loss of 0.1% per pass through the fuel-recycle plant
was applied. J

/ I , , -
Teble 12. Economic Ground Rules Used in
- Obtaining Fuel-Cycle Costs -

 

~Reactor power, Mw(e) - 1000
Thermal efficiency, ® b5
Load factor o 0.8
Cost assumptions _ _ _ .

Value of 22y and z3':3Pa $/g s 14
“Value -of - 5U $/g o 12
.~ Value of thorium, /kg 1
- Value of carrier salt, $/kg 26

Capital charge, %/yr
- Private financing = - n , 5
_Depreciesting capitel = 12

e Nondepreciating capital 10

- Public financing | Ll
~ Depreciating capital L T
Non&epreciating capital - -5

_fiProcessing cost-' given by curves{j;;
- in Fig. 11, plus cost given by
Eq. (1), where applicable.

 

 
 

 

Ly

.. The effective behavior used in the fuel-cycle-performance celcula-
tions for the various fission products was that given in Teblée 13. The
gas-stripping system is provided to remove fission- product gases from the
fuel salt. In the calculstions reported here, the **°Xe poison fraction
was assxmed to be 0 005.

'.I?a'ble 13 Behavior of Fission Products -
1n MSBR Systems

 

Behav:lor : " Fission Products

 

'Elements present as gases, assumed to 'be S Kr, Xe_ S
removed by gas stripping (e poison : o
fraction of 0.005 was applied)

- ' Elements that form steble meta]_lic colloids ; " Ru, Rh, Pd, \Ag, In
removed by fuel processing : T s
| Elements that form either stable fluor’ides' o Se, Br, N‘b, Mo, Tc,
_or steble metallic. colloids ; removed by - Te ’ I
- fuel processing | _ ' |
Elements that form steble fluorides less - ~ sr, Y, Be, la, Ce,
volatile than LiF; separated by vacuum o Pr, Nd. Pm, Sm,
distillation ‘Bu, G4, Tb
Elements that are not separated from the Rb, Cd,..Sn, Cs, Zr .

carrier salt; removed only by salt discard

 

The control of corrosion products in molten-salt fuels does not
appear to be & significant problem, so the effect of corrosion products
vas neglected in the nuclear' calculations. The corrosion rate of Hastel-
loy N in molten salte ie very low;- in addition, the fuel-processing
operations can control corrosion-pmduct buildup in the fuel.

' The importent paremeters describing the MSBR and MSBR(Pa) designs
are given in Teble 8. Many of the parameters were fixed by the ground
rules for the evaluastion or by engineering-design factors that include
the thermal efficiency, plent factor, capital charge rate, maximum fuel
velocity, size of fuel tubes, processing costs, fissile-loss rate, a&nd
the out-of-core fuel inventory. The parameters optimized in the fuel-
cycle calculations were the reactor dimensions; power density, core =
composition (including the carbon-to-uranium and thorium-to-uranium ratios),
end processing rates, _

{
™ C h

”-

»

b5

Nuclear Performance and Fuel-Cycle Cost

The general results of the nuclear calculations are given in Table 8;
the neutron-balance results are given in Table 1%, The basic reactor
design has the advantage of zero neutron losses to structural materials
in the core other than the moderator. Except for the loss of delayed
neutrons in the external fuel circuit, there is almost no neutron leskage
from the reactor because of the thick'blanket. The neutron losses to
fission products ere low becsuse of the low cycle times associated with
fission-product removal.f

The components of the fuel-cycle cost for the MSBR(Pa) and the MSBR
are sumarized in Table 15. The main components are the fissile inventory
and processing costs. The inventory costs are rather rigid for a given
reactor design, since they are largely determined by the external fuel
volume. The processing costs &are & function of the processing-cycle
times, one of the chief parameters optimized in this study. As shown by
the results in Tables 8 and 15 the ability to remove protactinium directly
from the blanket stream has & marked effect on the fuel yield end lowers
the fuel-cycle cost by sbout 0.1 mill/kvhr(e). This is due primarily to
the decrease in neutron absorptions by protactinium when this nuclide is
removed from the core and blanket regions.

In obtaining the reactor design conditions, the optimization pro-
cedure considered both fuel yield and fuel-cycle cost as criteria of
performance. The corresponding fuel-cycle performance is shown in Fig.
12, which gives the minimum fuel-cycle cost as a function of fuel-yield
rate based on privately finsnced plants end & plant factor of O. 8. The"

‘design conditions for the MSER(Pa) and MSBR concepts correspond to the

designated points in Fig. 12.

Power-Production Cost and Fuel-Utilization
Characteristics

The powerfproduction costs are based on the capital costs given
ebove, operation and maintenance charges, and fuel-cycle costs. Table 16
surmarizes the power-production cost and the fuel-utilization character-
istics of the MSBR(Pa) and MSBR plants. Both concepts produce power at
low cost and have good fuel-utilization characteristics. In terms of fuel
utilization, the MSBR(Pa) concept is comparable to a fast breeder reactor
with & specific inventory of 3 kg of fissile material per megawatt of’
electricity produced and & dofibling time of - 9 years, while the MSER plant
1ls comparable to the same fast breeder'with a donbling time of 12 years.

Alternatireé:to:the?Reference Design

- The MSBR and.MSBR(Pa) reference design represents extrapolation to
a large scale of technology that'has ‘been mostly demonstrated on a.mich
smaller scale. The major uncertainty is whether the graphite fuel cells
will have an economical life in the high fast neutron flux in the core.

 
 

 

 

Teble 4. Neutron Balences for the MSER(Pa) and the MSBR Design Conditions

 

 

 

 

 

 

 

 

 

'MSBR(Pa) - * MSBR
Neut.rons per Fissile. Absoxption Neutrons per Fissile Absorption -
Marerial Total gfigfifig . Feutrons  Total g:ggi:fig Neutrons
Absorbed Fission - Produced. . Absorbed Fission Produced.
3327n 0.9970 0.0025 0.0058 0.9710 -~ 0.0025 . 0.0059
..333pg 0.0003 o , 0.009 - - .

- @s3g o 0.9247  0.8213 2.0541  0.9119  0.8090  2.0233
334y | 0.0819 0.0003 0.0008 0.0936  0.000k 0.0010
sy 0.0753 0.0607  0.14Th 0.0881 < 0.0708  0.1721
ey ~ 0.008% . 0.0001 0.0001  0.0115  0.0001 10,0001
337Rp - 0.0009 0.001% 8 A
238y 0.0005 ' 0.0009 o
‘Carrier salt . 0.0647 . 0.0186 0.0623  0.0185

(except ®Li) ' -
811 0.0025 ~ 0.0030
- Graphite 0.0323 0.0300
136%e - 0.0050 : - 0.0050
149gm 0.0068 0.0069
181gy ~ 0.0017 0.0018
Other fission 0.0185 0.0196
products '
‘Delayed neutrons 0.0049 . 0.0050 -
‘lost? ‘ : :
Leskage® 0.0012 0.0012 -~ |
Total -~ 2.2268 . 0.8849 2.2268 2.2209  0.8828 2.2209

 

a?l)elta.:,red neutrons emitted outside core.
bLetsl.kage, Including neutrons absorbed in reflector.

o

.C
“\ -
Tflblg 15._-Fuel-cycle Cost for MSBR(Pa) and MSER Plants™’®

 

MSBR(Pa) Cost (mill/kwhr)

MSBR Cost [mill/kwhr(e)]

 

Fuel

 

 

_Stream.‘ g:::iié Subt0t51  g@ifi% 7Si32;m gii:iié Subtotal ggifi%
Fissile invéhtbryb, 10.1198 0.0208 0{1&13 0.1247 0.032Y4 0.1571
Fertile 1nvénto:y g o;dooQ - 0.0179 ,0.0179' 0.0459 o,ou59
Salt inventory .; | Qflo;56. 6.0226 - 0.0396 :Ao.015u' 0.0580 '0,073h |
Totalkinventory o / | | 0.20 | o 0.28
Fertile replacementl_.0.0000 '0.0041 0.0041 N | 0.0185 0.0185
Salt repiacement | '0;0636 0.0035  0.0671 - 0.0565 0.0217  0.0782
: Total replacefiént .. | 0.07 | | 0.10
Processing - 10.1295 70.0637 | 0.1932‘ 0.1223-' 0.0440 0.1663 |
Total processing" | 0.19 | - 0.17
Production. credit (0.10) (0.07)
Net fuel-cycle cost - 0.36 0.48

 

Based on investor-owned power plant and 0.80 plant factor.
Including 233pg ,. 233U and 225y,

Revised

4

 
 

 

ORNL-DWG 66-Th56A

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0.6

0.5
5 MSBR DESIGN POINT
3 I
'El 0.4
g MSBR(Pa) DESIGN POINT 67

1

g M
E 0.3 |

o;a ,

2 3 5 6 T - 8

FUEL YIELD (%/yr)

Fig. 12. Variation of Fuel-Cycle Cost with Fuel Yield in MSER and

MSER(Pa) Concepts.

9
r-‘:jn

4

49

, Table 16, Power-Production Cost and Fuel-Utilization Characteristics
| of the MSBR(Pa) end the MSBR Plants®’®

 

MBBR(P&)

 

mills/kwhr(e)

 

MSER
Specific fissile inventony, 0.12 0.81
- kg/mv(e) |
Specific fertile inventony, 101 260
ke/Mi(e) |
Breeding ratio 1.07 1.05
Fuel-yield rate, %/yr 7.5 b5
"Fuel doubling time,b years‘ 13.0 22,0
Power doubling time, yea:s_ - 9.5 | 15.0
Private Public .- Privete Public
. Finencing - Financing Financing Financing
Cepital charges, mills/kwhr(e) 195 1.10 1.95 1.10
Operating and maintenance cost, -  0.34 ©0.34 0.34 0.34
mill/kwhr(e) : |
Fuel-cyele cost,drmill/kwhr(e) . 0.36 0.21 0.48 0.30
i Powereproduction cost Lj-2.7 - LT 2-8:‘i5"‘ 1.8

 

®pased on lOOOéMw(e) plant end & 0.8 load factor.- Private. financing con-

Inverse of the fuel-yield rate. ) : :
Capability based on continuous investment of -the net bred fuel in new re-

' actors- equal to the reactor fuel doubling time multiplied by 0.693.

;siders e capital charge rate of 12%4 ‘yr for depreciating capital and of 10% r for
- nondepreciating capitel; public financing considers a capital charge rate of
17¢/yr for depreciating capital and’ 5$[yr for nondepreciating c&pital.

dCosts of on-site integrated processing plant included in this value.

Revised.'

 
 

 

 

 

50

This, :in turn, is relsted to the cost in equipment, effort, and downtime
to do maintenance of the highly radioactive core and other components

in the -réactor primary systems. Several alternatives to the reference
design have been proposed and they are primarily concerned with making
these problems lese difficult and in some instances with generally im-
proving the performance of the breeders. These alternatives and the
extent to which they should be included in the program of development
of large power breeder stations are discussed below. ,

Moduler Designs

_ The reference design has four fuel circulits and four‘blanket circuits
operating off one reactor vessel in order to produce 1000 Mw(e). One

- coolant circuit is provided for each fuel eand blanket circuit. If a

graphite tube in the core were to fail or & pump in the primary system

were to stop or & tube in & primary beat exchanger were to fail, the

entire plant would have to be shut down until the fault was repaired.

We believe the components can be mede-relieble enough so thet such shut-

downs will be infrequent, but they will heppen. . ,

As an alternative,'a modular design was evolved vith the objective
of providing assurance of high plent eavailsbility. Eech primary circuit
of the reference design and its secondary circuits were connected to &
separate reactor vessel to provide four 556-Mw(th) reactor modules. The
modules were installed in separate cells so that one could be repaired
vhile the others were operating. The lsyout is shown in plan and ele-
‘vation in Figs. 13 end 1k. ‘ ’ -

Although the modular design has four reactor vessels, they are
smaller than the reference vessel. The average power density in the
fuel gsalt and in the core are the same &s in the reference reactor; the
reactor vessel for each module is &bout 12 £t in diameter by 15 ft high
as compared with 14 f1 diam by 19 ft high for the reference design. Most
of the rest of the equipment in the two types of plents is the same, &nd
the plents are of very.nearly the same gize. The increase in total cost
of the moduleyr plant over the reference plant would be ebout L%; there
is no significant difference in breeding performance or in cost of the
power produced.

The reference design and the moduler deslgn described ebove operate
at the same high power density in the core end the graphite is subjected
to & high dose of demaging neutrons in a few years--10 neutrons/cm
(max) in four to six full-power years depending on the amount of flux
" flattening that can be achieved. This dose is & factor of 4 higher then
has been achieved to date in in-pile. testing, and having to replace the
graphite every 5 years is ‘estimated to increase the power cost by 0:05
to 0.1 mill/kvhr. Although there is considerable confidence that graphite
can be developed to perform satisfactorily to even greater doses, several
years of irradistion in the HFIR and in EBR-II or other fast test reactors
is required to provide & firm basis for this confidence. .
 
 
  
 
  
  
  
 
  
  

REHEATERS —

STEAM~ '
GENERATORS—-

INSULATION —-

COOLANT . |

 

PUMP -

DUMP TANKS

© 2w SALT —
BLANKET —
FUEL —-

- BLANKET Hx.

AND PUMP — | ©

PRIMARY Hx
AND PUMP

REACTOR —

s
v

o xd

~ STEAM PIPING - p——2¢-¢ B

L .

‘,.‘ et e

< TRARE GAS REMOVAL

.

OO0 O
FUEL
PROCESSING '.f :

e
T»

% W gy e

45T+ 0" — ‘
o - P

8-0° =— 22-0'

Fig. 13. Molten-Salt Breeder Reactor Plan of Modular Units.

' ( '

: SALT
4 | SToRAGE
-b
ECONTA
INATION &
RAGE
o
HOT CELLS

 

TG
 

  
 
  
  
 
 
  

" CONTROL ROD DRIVE

 

FUEL AND BLANKET
PUMP DRIVE MOTORS

 

OOLANT SALT .PUMPS
STEAM GENERATORS
'REACTOR-
REHEATERS
BLANKET HEAT -
EXCHANGER |
=
60-0" =
=3 STEAM
o ) . PIPING
PRIMARY HEAT - = -
EXCHANGER o

 

' . ] . " N ™ . -
A L e e e Y . SRR : i : e

 

. Lo ot T . .
4‘ a ¥ !

Fig.' 14. Elevation of Modular Units.

 
"

"

23

For these reasons the first molten-salt breeder reactors are likely
to be operated at lower power densities where an acceptable core life is
more easily assured, so considerable atiention is being given to a modular
plant in which the average power density in the core is L0 kw/liter--
half the power density in the core of the reference design. Again the
only significant physical change in the plant is in the size of the core
and the reactor vessel. The reactor vessels become about 13 ft in diam-
eter by 17 £t high; the breeding ratio remains about the same, but the
yield decreases; the capital cost would be sbout &% higher than for the

reference plant.  Some characteristics of modular plants with full and

half power density in- the}core, with and without. protactinium removal,
are shown in Tables 17 and 18. The plant factor is 0.8 as for the refer-
ence design, no credit being taken for being able to maintain a higher
plant factor. .

Whether the modular design represents a more attractive or a less
attractive alternative to the reference design deperids on the outlook of
each designer end operator. The modules can be made larger than 556 Mw(th)
if desired, the capacity depending on the fraction of plant the operator
is willing to have shut down for repair on short notice. No special devel-
opment is required for the modular design. It should receive continued
attention as design studies are made. -Construction of a plant of the
size of one module could be & desirable step in the development of large
power breeder stations.

Mixed-Fuel Reactor

In the reference design, graphite cells or tubes with. graphite-to—
metal Joints on one end are used to keep the fuel and blanket salts from
mixing in the reactor vessel. The major feasibility question in the
design is whether the damage to the graphite by the high flux of fast
neutrons will cause the cells to crack or break in less than the three
to five years required for replecement to be economical.

An alternative to this'type of reactor is one in which both thorium
and uranium are contained in the fuel salt which flows through channels
in graphite bars much as-it does in the MSRE. In order for the reactor

‘to be a breeder the core would have to be surrounded by & blanket as

shown in Fig. 15. The wall separating the core and blanket would be

‘Hastelloy N, niobium, or molybdenum, 1/8 to'1/k in. thick. Whether &
- satisfactory core tank can be developed ‘is the major feasibillty question
'of this reactor. : ,

The breeding performance of such a reactor ig shown in Table 19..

: The specific inventory and the douhling time can be attractively low.
- MaJor requirements are that - satisfactory processes be invented to sepa-
rate protactinium continuously from uranium end thorium in-the fuel

stream and to separate thorium from fission products. The demands on
fuel processing for this reactor are considerably greater than those

imposed by the reference ‘MSBR.

 

 

 
 

54

 Tablé 17. . Design Values for Modular Plants

 

' Full Power Density = Half Power Density

 

 

With Without . With - Without
Pa Removal Pa Removal ~ Pa Removal  Pa Removal

 

 

Power, Mw .
Thermal - per module 556 556 556 556
- - total - - 2223 2225 = 2223 2225
Electrical - Total : © 1000 1000 1000 1000
Core o * : : o P
Diameter, ft . 6.34 : 6.34 8 , 8
Height, ft 8 8 : 10 10
Number of graphite fuel 210 210 336 - 336
tubes ' ' ‘ , _
Volume, ft° 253 - 253 503 505
Volume fractions o ' - oD .
Fuel salt . 0.164 0.164 0.165. - - 0.165
Fertile salt 0.05 0.055 0.06 . 0.06
Graphite 0.786 0.781 0.775 0.775 ]
Avez:ge neutron fluxes, . S e :
n sec '
Thermal x 10'* : 6.56 5.62 B3 bl 3.3
Fast over 100 kv - 291 2.90 : 1.48 1.48
! x 10t4 S
Average power density,
kw/liter , _ , _
Gross 78 78 , 39 -39
Fuel salt k75 475 237 237

Average fuel salt
temperatures, °F _ o ‘ S
In 1000 1000 1000 1000

Out . ‘ 1300 1300 1300 - 1300
Fuel salt flow, ft2/sec 25 25 25 , 25
Blanket
Thickness, ft - , - , ,
Axial , 1.25 1.2 1.25 - 1.25
- Radial 2 2.32 1.5 , 1.5
Average blenket salt ’
' temperatures, °F , :
In 1150 1150 1150 1150
Out - 1250 1250 . 1250 .. 1250
Blanket salt flow, 0.2 1.2 0.2 1.8
sec - , . :
Volume fractions
Blanket salt 0.65 0.T1h4 60 60
Graphite ' - 0.35 . 0.286 ? 40 - o
Reflector thickness, in. 6 6 6 6 .,
»

22

Table 17. (Continued)

 

Full Power Density Half Power Density

 

 

. With Without . With Without
Pa Removal Pa Removal Pa Removal

»n o

b

Pa Removal

 

Reactor vessel
dimensions, ft

12.00

Diameter 11.4 12 12.00
Height ~13 ~13 ~LT7 ~L7
Salt Compositions, |
mole, %
Fuel . ' _
LiF 63.5 63.5 63.5 63.5
Blanket ' . o
LiF - 71 71 71 71
BeFs 2 2 2 2
ThFy a7 27 27 27
System-Inv%ntories o ' ,
Fuel salt, Pt° 169 169 229 229
Blanket salt, f£t° 532 1063 565 973
Fissile material, kg 175 217 218 253
Fertile material, W 81 43 5
1000 kg
Processing Data - Full Plant :
Fuel stream _ ’ -
Cycle time, days C30.4 34.5 50 - 50
Rate, £t3/day 20.8 - 18.4 17.6 17.6
Blanket streem ‘
Fluoride volatility
~ Cycle time, _
days . | W53 3T 50 50
 Rate, ft?/day 46 112 Bl Lk 76.4
Protactinium removal Ll P :
- Cycle time, days O.k2 - -- 0.k2 --
Rate, £t°/day ~ 5112 -- 5360 -
 Fet breeding ratio 1.06 1.05 1.07 1.05
'Specific inventory, 070 0.87 0.87 - 1.01
kg fissile/Mw(e) T R -
Specific pover, 3.2 2.6 2.6 2.2
Mw(th) kg fissile S T _ o
Fuel yield,_%[year | | - 6.8 4.6 6.0 3.9
Fuel doubling time, year 15 22 17 26
Reactor doubling time, yr 10 15 12 18

 

 

 

 
 

 

 

Table 18. Fuel-Cycle Costs from Modular Plants

~

 

Full Power Denéity

Half Power Density_

 

 

 

With ~ Without With Without
Pa Removal  Pa Removal Pa. Removal Pa Removal
Fissile Invefitory
Fuel Stream 0.1160 - 10.1300 0.1498 0.1524
Fertile Stream 0.0206 0.0397 0.0208 0.0458
Subtotal 0.1366 0.1697 0.1706 0.1982
‘Fertile Inventory 0.0287 0.057h 0.0305 ' 0.0525
Carrier Salt 0.0514 0.0878 0.0588 0.0868
Total Inventory 0.2167 0.3149 0.2599 = - - 0.3375
Salt Replacement 7
Fissile Stream 6.0868 0.0764 0.0732 - 0.0732
Fertile Stream 0.0069 0.0169 0.0067 0.0115
Subtotal 0.0937 0.0933 0.0799 0.0847 -
Fertile Replacement 0.0068 0.0146 0.0066 0.0104
Total Replacement  0.1005 0.1079 0.0865 10.0951
Processing
| Fissile Stream 0.1279 0.1216 0.1195 - 0.1195
Fertile Stream 0.0681 0.0368 0.0671 . 0.0316
Total Processing 0.1960 0.1584 0.1866 0.1511
Production Credit 0.0920 0.0760 0.1021 0.0766
Net Fuel Cycle Cost ' Q,ha , 0.51 ° ¢ 0.43 0.51.

 
»

n

i

 EXCHANGER |

57

- -FUEL PUMP MOTCR T eol . ROC DRVE s PUMP MOTOR

HEAT
EXCHANGER

FUEL PUMP

PRIMARY
HEAT

CELL
HEATERS

 

|
|
W ;._‘-..' '..:,.,7.,9_',-.',,7 _-.',--_..;. ._,.:.:- , - |
Fig. 15. Mixed-Fu
|
|
 

58

 

Table 19. Some Performance Data for Mixed-Fuel Reactor ,\fi%
Core size, ft | | | 10 diem x 15 high
Pover density in fuel, kw(th)/liter 360
" Fuel composition, mole % | 66 LiF-25 BeFo-
| : | o . 8,7 ThF¢—0.3 UFg
Specific power, Mw(th)/kg 2%y 3.2 |
Specific inventory, kg =3U/Mw(e) 0.68 | | |
Breeding ratio | o 1.06
 Yield, % per annum ' T.2
Fuel cycle cost, mills/kwhr(e) ‘ 0.33

 

, aAssumes'that processing is no more complicated or'expenSive' ’
than for reference MSBR.

This alternative is ettractive if serious problems are encountered .
with the graphite tubes of the reference design, but substitutes problems \ .
of & metal core tank and more difficult reprocessing. The neutron ab- '
sorption in the metal core tank increases with decreasing core size, so z
the breeding performance would suffer if & modular design were used and >
the reactor were made smaller to keep the specific inventory low. Work | '
on the mixed-fuel reactor should be limited to leboratory studies (or
observation of other groups! studies) of the effects of radiation on the
high-temperature properties of potential core-tank materisls, the compati-
bility of those materials with fluoride salts and graphite, and methods
of processing the fuel. If the results in the main line program indicsate
that the grephite cells are unlikely to perform satisfectorily in the
reference design, the development should be shifted to this mixed-fuel
alternative. The reactors are so similer that most of the work done on
the reference breeder would be appliceble to this alternative.

A

Direct-Contact Cooling with Molten Lead -

The reference-design MSBR has three volumes of fuel outside the core
in heat exchanger, piping, plenum chambers, etc., for each volume of fuel
in the core. Studies indicate that the fuel volume could be reduced to
gbout one volume outside the core for each volume in the core if the fuel
salt were circulated and cooled by direct contact with molten lead. The
-lead would be pumped into & jet at the lower end of each fuel tube. Salt
and lead would mix in the jJet and be separated at the outlet, The salt
would return directly through the graphite cells to the core and the lead
would be pumped either through intermediate heat exchangers or directly to
the steam.generators.

_ This system has seversal advantages. Ideally the specific inventory ./
could be reduced to 0.3 to 0. h kg of =3y per ’megawatt (électrical) and |
»n

29

the doubling time to 5 or 6 years. Relatively inexpensive lead would be
substituted for some of the lithium and beryllium fluorides. The lead

- pumps and heat exchangers could be arranged for maintenance of individual

units with the remainder of the plent operating. Some parts of the plant
should be considerably simplified.

‘There are some uncertainties also. Thermodynamics data indicate that
lead, fuel and blanket salts, graphite, and refractory metals such as
niobium and molybdenum alloys should be compatible. Preliminary tests
indicate that this is true and that the much less expensive iron-chromium
alloys might be used in the main lead systems. ‘However, the materials.
problems are almost unexplored; little is known of the effects of radi-
ation or fission products or of the ease of separating lead and salt.

' The lead~-cooled reactor represents an almost completely new technology
that cannot presently be given & good evaluation. Work on the basic
chemical, engineering, and materials problems of the system should be pur-
sued to make a good evaluation possible within three or four years. If
direct-contact cooling proved to be practical, its adoption could produce
impressive improvements in the performance of the thermal breeders and
could point the way to the use of molten-salt fuels in fast breeders.

PROGRAM FOR DEVELOPMENT OFZMOHEEN-SALT THERMAL BREEDER POWER PLANTS

We believe the information in the section on fuel utilization strongly
indicates the need for the U.S. to be able to build 1000-Mw(e) or larger
power breeder stations of high performance by sbout 1980, so they could be
built at a rate near 50,000 Mw(e) per year by about 1990. The development
program for a molten-salt thermal breeder should be aimed. directly at that

goal. This requires an aggressive program, carefully plenned and exe-

cuted and supported by firm intentions to carry it to completion unless
developments along the way show that the technical or economic goals
cannot be met. - S

- Steps in the Development”

The technology &s it presently exists is embodled in the Molten-Salt

. Reactor Experiment. The reactor is a one-region, one-fluid reactor. It

operates at 1200°F but at 7.5 Mw(th), so the power density is low. Some
exploratory tests, however, indicate that the fuel salts and the major

- structural materials--graphite ‘end Hastelloy N--should be: compatible at

pover densities far above the maximum in the reference breeder design.

* "The MSRE plant includes some provision for fuel processing and for mainQ
- tenance of radioactive equipment but much less than w111 be needed in
' . a power breeder plant.q,so,;. o . : -

_ Successful operation of the MSRE is providing an. essential base for
proceeding with larger reactors, but a true breeder pilot plant--a Molten-
Salt Breeder Experiment--should be operated before building a prototype
povwer breeder plant. The MSBE should include the essential features and

 
 

 

60

satisfy all the technical criteries of the reference design, but it should
be ebout as small a plant as will meet these requirements. According to
preliminary studies, the power would be 100 to 150 Mw(th). The experiment
‘would demonstrate &ll the basic equipment and processes under the most '
severe conditions of the large plants; lts essential purpose would be to
produce information rather than electricity.

A prototype pewer breeder station would follow the MSBE. The size
“would be 250 to 500 Mw(e), one module of the modular design described
above. A full-scale plant could then be obtained by adding modules to
the prototype plant or by building a plant of the reference design with
heat transfer circuits of the size developed for the prototype

Plans are discussed here and in relsted reports for designing, devel- ,
oping, and building the MSBE. They are aimed at having the experiment in
operation as soon as is consistent with resolving all basic prdblems
‘before beginning construction and major procurement for the plant. Detailed
design of the plent and research and development for all the parts proceed
concurrently. Design in detail is essential for identifying all the devel-
opment problems, and much of the development for & fluid fuel reactor con-
sists of building, testing, and modifying the equipment that has been
designed so that it will perform satisfactorily in the reactor.

Ruclear operation of the MSBE would begin in FY 197>. A prototype
could be in operation by 1980, and its construction would bring into
being the capability for building full-scale plants. This cepability
.could then be expanded according to the needs of the time. We have not
included & more detailed schedule or a projection of the development
costs for the prototype or for plants beyond the prototype. If the
MSBE fulfills its purpose, the development would consist largely of
building and testing larger equipment and improving on demonstrated
processes. The rate and manner in vhich the work on larger reactors
would proceed and the distribution of expenditures between government
and industry are uncertain and are completely out of our control. We
therefore have limited our projections to the essential step in making
this further development feasible and ettractive to the equipment industry
and the utilities.

Precent Status of the Technolog!_- MSRE_

The present status of the technology ie best described in terms of -
the MSRE and some supplementary information. The MSRE is & molten-salt-
fueled thermal reactor that produces heat at a rate of 7.5 Mw(th) while
operating at sbout 1200°F. - The purpose of the reactor is to provide &
demonstration of the technology end a facility for investigating the
compatibility of fuels and materials and the engineering features of ,
molten-salt reactors. The design conditions ere shown in the flow dia-

- grem in Fig. 16, and the general arrangement of the plant is ghown in
Fig. 17.
 

 

 

ARSOLUTE
FULTERS

 

 

 

ORML-DWE £5-H410

 

 

 

 

 

 

SAMPLER- o PUMP
ENRICHER - SAMPLER

 

 

 

 

 

 

10159F
e 850 GPM.
1210 :
OVERFLOW TANK .
LU I S : ' o | . ARFLOW: 200,000 cfm
. 1200 GPM. _ X . o ‘
REACTOR N : w073 F _ :
. VESSEL T - “FREEZE FLANGE (TYP} . " ' : —
a - ! ‘ pi——
FREEZE VALVE (TYR)
/ : RADIATOR
A

19

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

TANK

 

 

 

 

 

 

 

 

 

 

Fig. 16. MSRE Flow Diagram.

 
 

 

 

 

ORNL-DWG $3-1209R

 

 

 

 

 

e T,
-
a’ ' IA

 

e
REMOTE MAINTENANGE ‘
i
REAGTOR CONTROL -*’L
ROOM
.’;.a "; ':; ’ ‘.
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I. REACTOR VESSEL

3 FUEL Pume
4 FREEZE FLANGE

 

6. COOLANT PUMP

2 MEAT EXCHANGER

S. THERMAL SHIELD

~ N
[—--_";“w‘ Co 4
_.4"’ .
¥
v
=7 \;;-... . )
& |
4
o 9
8
7. RADIATOR \
8. COOLANT DRAIN TANK
9, FANS o
10. DRAIN TANKS
1L FLUSH TANK
2 CONTAINMENT VESSEL
13. FREEZE VALVE

Fig. 17. General’Arrangement of MSRE.

29

 
»

"

63

The fuel for the MSRE is 65% "LiF-29.1% BeFz5% ZrFe=0.9% UFq.¥
Except for the small amount of ZrFs and the higher'UF4 concentration, it
is the fuel for the core of the reference breeder.

In the reactor primary system.the fuel salt is recirculated by a
sump-type centrifugal pump through a shell and U-tube heat exchanger and
the reactor vessel. The flow rate is about 1250 gpm. The MSRE normally
operates at about 7.5 Mw thermal and at that power level fuel enters the
reactor at 1168°F. and leaves at 1210°F. The base pressure in the system
is 5 psig in the helium cover gas over the free surface of salt in the

‘pump bowl. The maximum pressure is about 55 psig at the outlet of the

pump.

The heet generated in the fuel salt as it passes through the reasctor
vessel is transferred in the heat exchanger to & molten-salt coolant con-
taining 666 TLiF and 34% BeFz. The coolant is circulated by means of &
second sump-type pump at a rate of 850 gpm through the heat exchanger,
normally entering at 1015°F and leaving at 1073°F, and through a radiator
where the heat is dissipated to the atmosphere. The base pressure in
this system is also 5 psig in the pump tank; the maximum.pressure, at
the discharge of the pump, is TO psig. .

‘ Drain tanks are. provided for storing the fuel and the coolant salts
at high temperature when the reactor is not operating. The salts drain
from the primary and secondary systems by gravity. They are transferred
between tanks or returned to the circulating systems by pressurizing the
drain tanks with helium. - _

The fission product gases krypton and xenon are removed -continuously
from the circulating fuel salt by spraying salt at a rate of 50 gpm into
the cover gas gbove the liquid level in the fuel pump tank. There they
transfer from the liquid to the gas phase and are swept out of the tank
by & small purge of helium. After e delay of about 1-1/2 hr in the
piping, this gas passes through water-cooled beds of activated carbon.
The krypton and xenon are delayed until all but the SSkr decay and then

. are diluted with air and discharged to. the atmosphere.. .

: Fuel and coolant systems are provided with equipment for teking
samples of the molten Balt through pipes attached to the pump tanks

while the reactor is operating at power. The fuel sampler is also used
for adding small amounts of fuel to the reactor while at power to com-
pensate for'burnup. ' ' T :

: Finally, the plant is provided with & simple processing facility for .
f:treating full T5-rt2 batches of fuel salt with hydrogen fluoride and fluo-
rine gases. . The hydrogen fluoride treatment is for removing oxide con-
temination from the salt as Hz0. The' fluorine treatment is the fluoride

volatility process for removing the ‘uranium as UFg. The equipment .
approaches the size required for batchwise processing of the’blanket of

" the 1000-Mw(e) reference reactor.

 

¥Percentages are in mole”%.

 

 
 

6k

. All the equipment in the MSRE thet contains salt is made of'Hastelloy
N. All of it was designed to be sble to operate at 1300°F. ‘The liquidus
temperature of fuel and coolant salts is near 850°F. It is desiresble to
keep the salts molten in the reactor systems and in the drain tanks, so
the major pieces of equipment ere instelled in electrical furnaces end
the piping is covered by electrical heaters and 1nsulation. '

: The reactor primary syetenn the fuel drain tenk systenb and some
auxiliaries become permanently radioactive during the first few hours of
operation at apprecisble power. Maintenance of this equipment end associ-
ated heaters, insulation, and services must be done remotely or semi-

- remotely by means of special tools. Tools have been developed for accom-
'plishing this maintenance of the MSRE eqnipment.

. The MSRE reactor vessel is. shown in Fig. 18. It is ebout 5 ft diam
by 8-1/2 £t high from the drein line at the bottom to the center of the
outlet nozzle. The wall thickness of the cylindrical section is 9/16
in.; the top and béttom heads are 1-1/8 in. thick. The core conteins
approximately 600 vertical graphite bars 2 in. square x 67 in. long.

Most  of the bars have grooves 1.2 in.-wide x 0.2 in. -deep machined slong
the full length of each face. The bars are instelled with the grooves
on sdjacent bars aligned to form channels 1.2 in., x O.k in. for the salt
to flow through the core. The graphite is & new type with high strength,
high density, and pore openings averaging sbout O.4 microns in diameter.
The salt does not wet the graphite and cannot penetrate through the small
pores unless the pressure is raised to 5 to 20 times the normal pressure
in the core. :

- - Preliminary testing of the MSRE was begun in July, and fuel and
coolent systems were heated for the first time for the prenuclear testing
in the fell of 1964. The reactor.was first critical in June 1965 end
reached its meximum power of about 7.5 Mw(th) in.June 1966. The accumu-
lated operating experience through May 12, 1967, is presented in Table
20. Major activities are shown &s & function of time in Figs. 19 and 20.

Table 20, Accumuleted Operating Experience with MSRE

 

Fuel system ' | :
Circuleting helium above 1000°F, hr 3465
Circulating salt ebove 1000°F, hr | 9050
Full thermal cycles, 100°F to 1200°F ‘ T

Coolent system ‘ : -

Circulating helium ebove 1000°F, hr 2125
~ Circulating salt above 1000°F, hr - 10,680
~ Full thermal cycles, 100°F to 1200°F - 6
Time critical, hr o 5790
Integrated pover, Mvhr thermal | 32,450

Effective full-power hours = | ) | hSlO B

 
 

n

_ SAMPLE ACCESS PORT—\_

N
FUEL OUTLET—"
CONTROL ROD THIMBLES -

CORE .
CENTERING GRID 4

GRAPHITE-MODERATORY —RY ¢
STRINGER —— |

FUEL INLET
 REACTOR CORE CAN-
REACTOR VESSEL-

. » -

ANTI-SWIRL VANES—" ——~

VESSEL DRAIN LINEY

 

 

 

RN

65

ORMNL-LA~ DWG lmfiu

 FLEXIBLE CONDUIT TO
~CONTROL ROD DRIVES

-COOLING AIR LINES

ACCESS PORT COOLING JACKETS

R

' 3 /aencron ACCESS PORT

R -‘J.‘ R g

AR

FLOW DISTRIBUTOR

 

 

 

 

    
  

by

2N

= B
ey

S

  

e

-

S MODERATOR
SUPPORT GRID

Fig; 18. Reactor Vessel.

 
 

INSPECTION AND

PRELIMINARY TESTING

FINISH
INSTALLATION

 

TEST AUX. SYSTEMS
TR

'OPERATOR TRAINING .
:

'LOAD SALT
'INTO DRAIN TANKS
COOLANT  FLUSH

SALT

LEAKTEST,
PURGE 8 HEAT

OF SALT SYSTEMS SALT SYSTEMS
T T e 277 )

ORNL-DWG 67-978

PRENUCLEAR TESTS
OF COMPLEJE SYSTEM

e N r

. FINAL PREPARATIONS
FOR POWERAOPERATION

 

INSTALL CORE SAMPLES
" INSPECT FUEL PUMP
HEAT-TREAT CORE VESSEL
‘ 27727

TEST SECONDARY
CONTAINMENT
T

INSTALL
CONTROL RODS

SAMPLER-ENRICHER

' . " FINISH VAPOR-COND. SYSTEM
OPERATOR

TEST TRANSFER, MODIFY CELL PENETRATIONS  ADJUST & MODIFY
FILL & DRAIN OPS. TRAINING REPLACE RADIATOR DOORS  RADIATOR ENCLOSURE
2222772 : Q27207022222 e/

LOAD U-235 : .
- | LOAD 8  IN ZERO-POWER LOW-POWER
CIRCULATE CIRCULATE ~  NUCLEAR {0-50 kw)
SALT = C & FL SALTS CARRIER _ EXPERIMENTS " EXPTS.

7l 222 2777700 08 2l l2r

-4

1 1 1 —1 i

——

1965
Fig. 19 MSRE Activities - July l964-—December 1965.

O

99

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o N b O ®

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

POWER MW (TH)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

INVESTIGATE - . 60-TO  REMOVE CORE SPECIMENS. . REPLACE  REPLACE  REMOVE
OFF - GAS PLUGGING,  FULL POWER. - REPLACE MAIN BLOWERS. -~ REMOVE . AIR LINE" BEARINGS =~ CORE
CHANGE CREPAR THAW FROZEN LINES. SALT PLUG. DISCONNECTS ON MAIN ~ SPECIMENS
_FILTERS AND VALVES. SAMPLER . TEST CONTAINMENT. - ~ CHECK =~ AND . BLOWER.  GENERAL
LOW-P - - - | REPLAGE OFF-GAS = CONTAINMENT, OFF-GAS FOLLOW  MAINTENANGCE
DYNAMICS < cHEGK . FWTER. . FumER. '¥xe |
TESTS. cou‘rmwam R = oL e ‘DECAY .
3 A1 B4 0 27/ B f [ ’//
W P A | | //4, ////i
o L ék A A 77 Z /// ///
4| Ll | oA ‘ W //’// . f’ //JZ A
SALT CIRCULATENG ';ES'gH',?; B Sl 1
L—-----"-----:1 i ' e m m‘-ma ‘ T P77 AN 0 N T T T A P77
Jo F‘ M .A-' M0y A s o0 N D  J F M A Mo
o - i966 e - “‘ s . 1967

Fig 20. MSRE Activities - J'a.nuary 1966—Ma.y 1967.

i9
 

 

68

(:‘}
K o "W

In most respects the reactor has performed exceptionally well. |
Analyses for corrosion products in the salt indicate that there has been
essentially no corrosion of the Hastelloy N by the selt. Inspection of
some parts of the fuel system confirmed that the corrosion was negligible
during about 1890 hours of circuleting salt in prenuclear and critical
tests., Samples removed from the core showed no attack on metal or =
graphite during the 2760 hours of suberitical and power operation from
December 1965 through July 1966. Analyses of the fuel salt for uranium
and reactivity balences indicate that the fuel has been completely stable.

. Although there have been problems with auxilieries and electrical
systems, few problems have been encountered with the major reactor systems,
The time to reach full power was extended several months by plugging of
small lines in the off-gas system that handles the helium and gaseous
fission products from the pump bowl. The reactor was shut down from mid- -
July to mid-October, 1966, by failure of the rotary elements of the
blovers in the heat rejJection system. After power operation was resumed
in October, it was interrupted in November and agsin briefly in January
for work on the off-gas line &nd on problems associeted with monitoring
of the reactor containment. In spite of these interruptions the reactor
vas critical T5% of the time--mostly at full power, the fuel system
operated 86% of the time, and the coolant system operated 100% of the
time from mid-October until the reactor was shut down in mid-Mey, 1967,
to remove graphite and metal specimens from the core. The major inci-
dents are discussed more fully below. ; '

e

The radiator housing is a large, insulated, electrically heated box
around the radiator coils and is required so the radiator can be kept hot
and the salt in it molten when the reactor is not producing fission heat.
The difficulties were in obteining proper operation of the doors and in
controlling leakage of hot air through Joints end through ducts for elec-
trical leads to prevent overheating of equipment outside the housing. -
Future molten-salt reactors are unlikely to have similar radiators, but
the experience will be helpful in designing better furnaces for other
equipment.

The off-gas system was designed for a small flow of gas, essentielly
free of solid or liquid serosols. Some difficulty was experienced with
micron-size particles of salts collecting in the tiny ports of the flow
- control valves, but much more difficulty was experienced after the resac-
tor began to operate at 1 Mw with organic solids and viscous orgenic
liguids collecting in the valves and et the entrance to the carbon beds.

The bearings on the fuel circulation pump are lubricated and parts
of the pump are cooled by oil. The oil is separated from the pump tenk
by & rotery seal. Provision is made for directing the normal seal leak-
age of 1 to 10 cc per day of oil to a waste tenk and preventing liquid
~ or vapor from coming in contact with the salt or cover gas in the pump tank.
Under speciel conditions, demonstrated in a pump test loop, this o0il can
leek through & gasketed seal in the pump presently in the MSRE and into o~
the pump tank where it vaporizes. The vapors mix with the helium purge \aj
stream eand flow into the off-gas system. The 0il has no effect on the
o

»

69

fuel salt but the organic materials polymerize in the off-gas system

“under the intense bete radiation of the gaseous fission products to form

the viscous liquids and solids that plugged the valves and the entrances
to the carbon beds. |

This problem has been reduced to & minor nuisance in the MSRE by
installing absolute filters for trapping solids and heavy liquids ahead
of the control valves. The leakage path has been eliminated in future
pumps by substituting & welded seal for the gasketed seal. Small amounts
of organic and inorganic vapors or aerosols are likely to be found in the
off-gas from future reactors, but they can be easily controlled by the use

~ of filters, traps, and absorbers._

: - The off-gas line was plugged once by frozen salt. This happened
vhen the pump bowl was accidentsally overfilled while the calibration of
the liquid-level indicators was being investigated. Salt was discharged
into some of the lines attached to the pump bowl and froze in the cold

- sections. Hbaters vere applied to the lines to remove most of the salt,
dbut it was necessary to open the off-gas line and break up.& small amount

of materisl in part of the line. Careful attention must be given to the
interface between hot systems and cold systems in the breeder designs.

: The maximum,power reached in the MSRE is 20 to 25% below the design
pover. It is limited by the heat transfer performance of the radiator,
but the overall heat transfer coefficient of the primary heat exchanger
is also less than had been calculated. In the case of the radiator the
air-side coefficient is low., While this indicates that better relation-
ships would be useful for calculating the air-side performance of such
devices, the designs for molten-salt breeder reactors do not contain
salt-to-air radiators. Recent data indicate that the equations used to
calculate the performance of the primary heat exchanger were adequate,
but that too high a value was used for the thermal conductivity of the
salt. This points to the. need for very good data on the properties of
salts for the breeder reactors.___ _

One day in July, 1966 when the reactor was runnlng at full power,

the power slowly decreased from T.5 Mw to about 5.5 Mw without action
- on the part of the oPerators,sand at the same time the air flow through .
,“the radietor decreased. Investigation soon showed that the reduction in
air flow hed resulted from the. disintegration of the rotary element on

one .of the two axial blowers that operate in parallel to pump air through
the radiator et a rate of asbout 200,000 £+t3/min, Although the blower was

 wrecked, the housing retained most of the fragments. Only some small .
- ,pleces vere blown through the radiator and they did no- damage

" The. rotary element on’ the other blower had a large crack 1n the hub,
s0 one blower and the rotary ‘element of the second had to be replaced.
Tt took &bout three months to dbtain,rotary elements with blades that

- would pass & thorough- exemination. The cause for the failures was never
~ fully esteblished. The blowers with new rotary elements have been

operated for about 8 months with the vibrations and bearing temperatures

. monitored carefully. One bearing on one blower has had to be replaced

 

 
 

 

T0

to keep the vibrations within specified limits. The rotery elements have \fiJ
been inspected periodically and show no signs of cracking. While this "
incident caused & long shutdown it is unrelated to molten-salt reactor - .
technology.

We stated above that the meChanicalxperformance of the MBRE salt
systems has been excellent, that there has been little or no corrosion
of the container metal and little or no reaction of the salt with the
graphite, and that the fuel salt has been completely steble. This is
the performance that the. component tests and several years of materisls
-work and chemlical development prior to the experiment had led us to \
expect. Aside from the experience with polymerization of organic materiels
in the off-gas system, the only unexpected behavior in the system has been
thet of fission products from.nidbium, atomic nunber U1, through tellu-
rium, atomie number 52 |

 These elements vere expected to be reduced to metals by the chromium
in the Hastelloy N end by the.trivelent urenium in the selt and to deposit
on metel surfaces in the reactor or to circulate as colloidal particles.
However, they were found in ‘considerable smounts on grephite as well as -
on metal specimens that were removed from the core of the reactor in
August 1966. Also there is some evidence of these materials in the gas ' _
phase gbove the salt in the pump bowl. In the higher valence states, s
.most of these elements form volatile fluorides, but the fluorides should
not be able to exist in equilibrium with the fuel sslt. The actual state
of these materiels in the MSRE may be exactly what the chemists expected; v
the deposits on the graphite samples may be thin films of metal particles;
and the materials in the gas phase may be serosols instead of volatile
fluoride compounds. More work is required to firmly esteblish the be-
havior of these elements in the MSRE end to relate this behavior to the
conditions of breeder reactors.,

In its performance to date the MSRE has fulfilled much of its
original purpose. Continued operation of the reactor now becomes important
in the investigetion of details of the technology, of long-term effects,
and of some aspects that were not included in the original plans.

The MSRE is the only large irradiation facility availsble or proposed
for use in the development of molten-salt reactors before the MSBE begins -
to operate. It is needed primerily for study of the chemistry of the fuel
salt end the materiasls. Continued investigation of the mechanism of depo-
sition of fission products on graphite end metal surfaces and of the’

. appearance in the gas phase of elements from niobium through tellurium is
essential to the design of molten-salt breeder reactors. This information
will be obtained through studies of the fuel salt, the off-gas from the
pump bowl, end. specimens of graphite end metal that are exposed in the

core and in the liquid end vepor phases in the pump bowl. The core of

the MSRE is the only place where large numbers of specimens can be accom-
modated for this purpose and also for determining the effects of irradi-
ation-on metels and grephite in a fuel-salt environment. Since the MSRE \ —~F
- operates at low power density, the effects of power density must be f O
determined in capsule &nd in circulating loop experiments in other reactors.
T1

By having these latter tests complement those in: the MSRE the number of
tests and the size and complexity of the test facilities should be con=-
siderably reduced. . :

Large breeder reactors will use 2330 as fuel and in the circulating
reactor the effective delayed neutron fraction will be reduced. to about
0.0013. This is much smaller then has been used in reactors to date
end has important safety and control implications. Plans are to fuel
the MSRE with 223U lete in FY-1968 and to investigete the stability of
the reactor vwhen operating with the small delayed neutron fraction,

This will be the first reactor fueled with 223U and good agreement
between the calculated and measured stability cheracteristics will give
confidence in the calculated stebility and safety characteristics of
the large breeders. ' _

While- the above experiments are in progress the longer operation of
the reactor will subject the equipment to additional exposure to radi-
ations and operation at high temperature. Effects observed and experience
with the equipment will provide data helpful in designing the MSBE and
in design studies for larger plants. Experience with the maintenance
and studies of radiation levels and the principal sources will apply
directly to the development of maintenance methods and equipment for
those reactors. , , : _

Advances in Technology Required for 8
High-Performance Thermal Breeder -

Advancing the technology of the MSRE to the level required to build
large, two-flnid two-region power breeders requires few, if any, major
inventions. It does require considerable research and development to
increase the depth of knowledge in ‘the entire field, to improve materisals
and processes, to make larger, better equipment, and to demonstrate a
much ‘higher performance in a combined reactor, processing, and pover
plant. ' : S L _

. The most important difference between the MSRE end the reference
breeder is the power density in the fuel. The maximum power density
in the fuel in the power breeder ie expected to be 600 to 1000 kw/liter,
‘a factor of 20 to 35 sbove the maximum in the MSRE. Results of short-
term in-pile tests of fuel salt and. grephite in metal capsules at 250
kv/liter and fuel salts in metel capsules at several thousand kilowatts
per liter indicate that the fuel is steble and compatible with the mate-
.rials at the high power density. This compatibility must be more thor-
- oughly established by tests of long duration under conditions proposed
for the breeder and, in some instances, under more severe ‘conditions.

A very important pert of this effort is to determine the distribution
of Pission products in the systems and in particular whether enough of
them deposit on the graphite to seriously effect the breeding potential
of the reactor.

 
 

 

 

T2

The two-region breeder makes use of graphite tubes or fuel cells to
keep the fuel salt from mixing with the blanket salt in the reactor core.
This graphite will be subjected to & maximum neutron dose of about 1022
nvt (E > 100 kev) in five years et the high power density in the center
of the core. The graphite bars in the MSRE have cracks that would pass
.salt, but with some additional development, tubes or fuel cells could.
almost ceértainly be made with the same low permeability to salt and free
from cracks. Whether they would survive the large rsdiation dose is
uncertain because no graphite has yet been irradiated-beyond sbout 3 x 1022
nvt. A more radistion-resistent graphite, possibly an isotropic material,
with equelly low permeability mey have to be obtained to get the desired
life. , :

The Hsstelloy N used in the MSRE hss excellent properties when un-
 irradiated, but the creep properties deteriorate under irradiation. This
behavior occurs in stainless steels and other alloys and is ceused by.
helium bubbles in the grain boundaries produced by thermal neutron irra-
diation of boron in the alloy. For the reactors to have long life, the
Hastelloy N must be improved to have better high-temperature properties
under irradistion. Research in progress indicates this cen be done, but
a satisfactory improvement must dbe demonstrated.with commercial materials.

The;vecuum distillation, protaotinium removal, and continuous vola-
tility processes for the fuel and blanket salts must be taken through the
laboratory &nd pilot plant stages.

. Equipment for the full-scale breeder plants and for any demonstration
plant will be considerably larger then that in the MSRE. Techniques
developed for building large equipment for other types of reactors will
have to be adapted to the needs of molten-salt reactors. Supercritical
steam generators, salt to steam reheaters, large pumps with long shafts
and molten-salt bearings and new concepts in cover-gas systems must be
developed for the reactors. A comtinuous fluorinator, & high-temperature
vacuun still, a liquid-metal to molten-salt extraction system and other
new devices are required for the fuel processing pleant. Equipment and
techniques must be developed for maintaining larger radioactive equipment
with greater facility. Development of remote welding end inspection of
radioactive systems is expected to be necessary.

All these developments must be combined and the new level of tech-
nology demonstrated in & breeder pilot plant. .

Criteria for the Molten-Selt Breeder Experiment
The MSBE should demonstrate all the basic technology of a large
molten-salt breeder reactor so that moderate scale-up and normal improve=-
ment of equipment end processes are all that is required to build lsrge
plants. The plant should be &s small and the power level &s low as is
consistent with making a complete demonstretion. Major criteria for the
plant are the following. .
a4

73

The average: core pover density in the fuel salt in the core should
be &t least the 470 w/liter of the MSER reference. esign.

Fuel, blanket, and coolant salts should be essentialxy those proposed
for use in the reference reactor. The uranium concentration may be
somewhat higher in the fuel salt in the experiment with the reference
concentration of thorium in the blenket but not so high as to cause '
the chemistry to be significantly different. A fuel of the reference
uranium concentration could be demonstrated by reducing the thorium
concentration in the blanket for the demonstration period.

The design of the plent should be similar to that proposed for a
large breeder and the components should be of a size and design that
can reasonebly be scaled up to meke components for & prototype. The
core should have graphite tubes or fuel cells with fuel salt in the
tubes and blanket salt around the tubes. Components probably should
be at least one—tenth the size of the components of the reference
design.

Reactor and coolant systems mnst be capable of operating with the
maximum.temperatures and temperature differences.

The reactor should be & breeder with high enough yield to demonstrate
breeding in & reasonable time. Suggested times are one full-power
year for the determination based on enalyses of core and blanket
fluids and weights: of fissile material fed to the core and removed
from the blenket and three to five years for a material balance over
the reactor and processing plant.

Methods for processing the fuel and blanket salts should be those
proposed for the reference breeder. Protactinium removal should be
included. Equipment for the processing plant should be of a size
that can be scaled up for the larger plant. Intermittent operation

- of the pilot plant would be acceptable to permit use of equipment of

larger size .

Maintenance methods and tools\should ‘represent major steps in devel-

-opment of equipment-for large power breeders. This probably requires

development of remote welding that might not otherwise be needed in

the pilot plant..

Supercritical steam should be generated in the- pilot plant and should

‘be used to produce electricity, This may require a special turbine,

| smaller than is normally built for use with supercritical steam.

o ;_” Results of some preliminary studies suggest thst e reactor with a
 power level of 100 to 150 Mw(th) would satisfy these criteria. Some
characteristics. of pilot plants of several sizes and power levels, but

with an average power density of 470 w/liter in the -core, are compared
with those of the reference design and one module of the modular alterna-
tive in Table 21. All the reactors use fuel cells of the same design,

 
 

Tsble 21. Comparison of Characteristics of Full-Scale and Pilot Plant Breeders.

 

Reference Modular

 

| Design Design MSBE Studies
.Power level, Mw(th) - | ) 2225 556 150 - 110 .44
L Mw(e) . 1000 250 70 50 20
" Core size | o - T -
" 'Diameter, ft_fi‘ 10 6.3 ka1 3.7 2.7
. Hedght, £t | ~ - 12.5- 8.0 5.1 46 3.
Blanket thickness, ft | - 2 2 2 2 - 3
Reactor vessel size - - | L | o \ IR ‘
‘Diameter, £t o I 14 | 12 9 - 8.7 9.7
'Height, £t _ . - 19 - 13 - 10 10 - 11
Fuel circulation rate, gpm I bk, 000 11,000 3000 - 2200« 900
Temperature rise, °F | | | 300 ~ 300 300 . 300 - 300
233y concentration in fuel salt, mole % 0,22 = 0.5 0.4 - 0.31 0.53
Thorium concentration in blanket salt, mole % 27 ' 27 et 2t 27
Fissile inventory, kg - 812 Lo 22T 120 Th ho
Core composition, volume fraction - i o - S o -
‘Fuel salt = N 17 17T 17T 18 15
Graphite i S CT6 78 8% 81 8
"Blanket salt ' ' o T 5 2 1r - 1
Blanket composition, volume fraction ‘ : o o | ' _ o
Blanket salt = o - 100 mn . 8 - 8 A
Pover density in fuel salt, kw/liter | ¥TO . b70 k0. M0 . 530 .
Specific pover, Mw(th)/g 2.7 | 2.6 1.2 1.5 1.1
Specific inventory, kg Mw(e) o ‘0.8 . 0.87 1.4 0 1.1 2.0
Breeding ratio o o 1,05  1.05 - 1.06 1.06 1.04
Fuel yield, % per year L ! - ks L5 2,5 3.1 .1.5
net production rate, kg/day - 0.13 0.033 0.0L 0.008 0.002
- Processing. rates, £t%/day S S S |
Fuel salt " 15 4.5 0.8 0.5 0.3
Blanket salt X - - 14h 28 T 4.3 5.4

‘ w

 

 

 

¢ ()m | ‘ | . | ta

o
 

W)

T

but the number and 1ength*vary_with core size. Moderator pieces around

 the fuel cells are modified to vary the fraction of blanket salt in the

core. The pilot plant would be expected to be a smaller version of the
modular design in having one fuel salt, one blanket salt, and one coolant-
salt circuit to remove the heat generated in the reactor. The comparison
suggests that a 100- to 150-Mw(th) reactor would satisfy the criteria.
For smaller reactors, -the fraction of blanket salt in the core becomes
impracticebly small, or the urenium concentration in the fuel salt unde-
sirably high unless the core is made drastically different from the
reference design._

SUMMARY OF PLANS, SCHEDULE, AND COSTS

-

Molten-Salt Breeder Experiment

The entire program centers sbout the breeder experiment. A proposed
schedule for the experiment is shown in Table 22. Conceptual design &nd
planning would begin immediately to provide the design basis for FY-1969
authorization of Title I and part of Title 2 design for & construction

- project. Authorization of construction would be requested for FY-1970.

Construction of buildings and services &nd procurement of major equipment

~ would begin in FY-19T1, this time being determined by the time required
for parts of the final design and for essential development work. No

construction or procurement would begin until all basic questions of
feasibility were satisfactorily resolved.  Prenuclear testing and check-
out of parts of the plsnt would begin in FY-l97h and +the plant would

reach full power in 1975.

The MSBE would be & complete power'breeder plent designed to operate
at 100 to 150 Mw(th) and to produce 40 to 60 Mw(e).. The experiment would
contain a reector and supercritical steam-generating plant, an electrical
generating and distributing plent, & fuel and blanket processing facility

- assoclated with the reactor, waste handling and storage facilities, and

all necessary maintenance equipment. - Preliminary estimates of the cost
of the experiment and the startup are presented in Table 23. The plant
costs represent & factor of more than two escalation of costs obtained by

scaling down to the experiment size the estimates for the lOOO-Mw(e) MSBR
'and the 250-Mw(e) module.r R

Training of operators, which is done in conjunction with the operation

of the Engineering Test Unit and the Fuel Processing Pilot Plant, and
,lfstartup costs were estimated on the basis of experience with the MSRE and
a8 variety of processing plants. Lo - -

Engineering Test Unit and.Fuel Processing Pilot Plant

As an important part of the development and testing of equipment ve

”plan +to build end operate a full-scele mockup of the reactor primary system,

coolant system, and fuel snd blanket processing facility.  Equipment
for this plant will be made directly from the early designs of equipment

 

 
 

 

Teble 22, Proposed Schedule for Molten-Salt Breeder Experiment

 

 

 

 

 

Fiscal Year 1968 . 1960 1970 1971 1972 1973 197h 1975
‘Conceptual design and plemning e | | |
Issue direction | oy
Design and inspection ) — : _4'
‘Construction of buildings and services \ ' . y R
Prbcfirement and ‘installation of equipment — ' ‘ ]

Prenuclear startup
Nuclear startup

 

 
- Teble 23. - Summary of Estime.‘bed Costs for Development, Cmstruction, and Stertup of the

Molten -Salt Breeder Experiment

 

Costs in Millions of Dollars

 

 

0.1

1.8 " '

125

 

1968 1969 1970 1971 1972 1973 197k 1975 Totals
Molten-Selt Breeder Ebcperinient o |
 Design and inspeetion | 0.8 1.0 20 2. 15 1.3 0.7 10 ;
Conatmction of build.il.ngs and. services - 0. 1.5 0.6 2. 5 40
Procurement and. installation of equipment 2._5' "8.0‘ 15 4.0 | 27, 5 i :
~ Operator Tre.ining a.nd. Sta.rtup of MSBE 0.2 02 02 1.k 40 'u,o*,:. - 10
Engineering Test Unit and Fuel Processing Pilot Plant | o - } o
Design and inspeetion L 03 0.5 0.5 - 02 0.2 0.1 1.6
Modification of building and services 0.2 03 | 0.5 )
Procurement end installation of equipment 0.5 %0 b2 0.5 9.2 s »
Preparation end operation 03 03 1.0 =28 LT 0.6 0 6.7 ; .
Development of Components a.nd Systems 1.3 -3;1 1.7 1.3 L 0.8: 0.6 | 0.4 Oh- : | 9.6
Instrumentation and C'ontrols Development 0.3 . 0.5 - 0.4 0.2 O.l 0.1 01 Ol "
Materials Development o ' 2.0 2.2 21 L6 09 0.5 o1 0.1 | 9.5
Chemical Research and Development | 1.2 1.6 2.1 2.2: - 2.0 1.8 1.y '_1-.3 g 13.6
' Fuel and Blanket Processing Development 1.0 2.3 "3‘.0 2.8 2.3 2.’:. 2.0 1.0 16.5 .
Maintenance Development 0.3 0.6 0.6 04 04 03 0.2 0.2 3.0
Physics Program 01 05 05 0.2 .02 02 02 0.2 2.1
Safety Prograanh ‘0..3 0.3 . 0.2 . 0.1 0.1 0.1 0.1 1.3

Lils
 

T8

for the MSBE and will be made of materisls being developed for use in the
finel plaent. The equipment will be arrenged in heated cells of the design
proposed for the MSBE but the cells will not have heavy concrete walls &nd
will be instelled in an existing building. _

Febrication of the equipment will provide manufacturers with'their
first experience in making reactor equipment of Hastelloy N &nd should-
result in much better equipment for the reactor. Operation of the plent
will provide a better test of the equipment, the methods of support, and
the furnaces than would individual tests. Maintenance procedures end
equipment will be tested there also. Operators for the MSBE will receive
much of their training in this test facility. Serious work on the test
plant is plenned to begin in the middle of FY-1968 with the goal of having
it in operation by the end of FY-19T71. Operation will end in FY-19Tk.

Development of Components and SystemsS

Much of the development and testing of components &nd systems will be
carried out in conjunction with the Engineering Test:Unit. In eddition
there will be extensive design, development, and loop testing of pumps
for the fuel and blanket systems and some work on the coolant pumps.
Relieble pumps are essential to long continuous operation of the reactor,
and the pumps for the MSER differ considersbly from those in use in the
MSRE. Other majJor activities include development of control rods e&nd
drives, & cover gas recirculetion system, mechanical velves for use in
salt, and parts of furnaces end speclal coolers. Flow tests will be made
in the ETU and in reactor core models. Heat transfer studies will be
made for the heat exchangers, the steam generator, and the reheater.
Minor testing will be done of components for the steam system and the
salt sampler, and the drain tank cooler systems developed for the MSRE
will be upgraded for use in the MSBE. Models of the pumps, the control
rods, and the cover gas &nd xenon stripping system will be operated,
solutions to other criticel problems will be demonstrated, and critical
parts of the heat transfer and flow tests will be completed in FY-1970.

Instrumentation and Controls Development?’

The instrumentation for the MSBE will depend heavily on the experi-
ence with the MSRE. Upgrading of some instruments wlll be necessary;
there will be considerable testing of the instrument components specified
for use in the MSBE. An ultrasonic flowmeter will be investigated for
measuring the flows of salt in the fuel, blanket, and coolant systems in
the reactor end in the ETU. Development of the control rods and drives
is included under the Component and Systems Development. The instrumen-
tation offers no barriers to the successful construction and operation of
the breeder experiment. :
”» C 3

¥

.t

79

Materials Developmente

 

‘Demonstration of a graphite satisfactory for the tubes for the core
of the reactor and & Hastelloy N with adequate high-temperature properties
under irradistion for making the equipment and piping are erucial items
in the development for the MSBE. The metals program includes modifying
the present Hastelloy N, testing the resistance to radiation effects, and

frdemonstrating that the lmproved alloy hes satisfactory corrosion resist-
' ance, weldability, fabricability, and compatibility ‘with graphite.

The graphite program includes determining the effects of very large
doses of fast neutrons on the properties of several promising graphites,
developing graphite in tubes with an acceptably high, resistance to radi-
ation effects and low permeability to salt and gaseous fission products,
and developing a satisfactory method for Joining the graphite to metals,
The program.is aimed at demonstrating before FY-19T71 that these problems

‘have adequate solutions. A strong continuing program is required in

support of the effort to provide all the Hastelloy equipment and & graph-
ite core for the MSBE. o

T Chemical Research and Development‘-’

Although the fuel salt for the MSBE 1s similar to the fuel used in
the MSRE end salts similer to the blanket salt have been used in experi-
ments, some studies must be done with salts of the actual compositions
proposed for the MSBE. The proposed coolent selt is new and must be
thoroughly tested. Details of the phase relationships will be obteined
in the vicinity of the Specified compositions. The physical and thermo-
dynamice properties and the behavior of oxides and oxyfluorides in the
saltS’will be studied in regions of interest to MSBE operation.

In-pile tests will he TN to establish the compatibility of salt,

~ graphite, end Hastelloy N through long exposures at high power density.

Good knowledge ‘of the distribution of the fission products between the
salt, graphite, and metal surfaces promises to he a very important result
of these experiments. _‘
_ |
Studies will be made of | protactinium end fission-product chemistry to

-provide & better chemical basis for the separations processes. Some work
will he done to improve the efficiency of the salt preparation processes.

Continuous knowledge of the composition of the salts, especially
the fuel salt, is desirable for running & liquid-fuel reactor. The most
direct way of obtaining this ‘information is through in-line enalysis of

"the salts. Effort will be spent on methods which have been partly devel-

oped under other programs and appear to be promising for making the

-analyses.

A favorsble fission-product distribution and good compatibility of

~ salts, graphite, and Hastelloy N et high power density are essential to

 
 

 

the success of the MSBR as'a breeder. e program is planned to pro-
vide definitive date by the end of FY-1 TO. '
Fuel end Blenket ProceSBing Developmenti®

The fuel and blanket process development involves converting the .
fluoride volatility process from batchwise to continuous operation end

taking the vacuum distillation end the protactinium removel processes from

the stage of demonstretion of basic phenomene in the laboratory to en
engineered plent. This includes developing flowsheets end equipment,
 determining effects of operating variebles, testing the processes in the

leboratory and pilot plants, and testing the final equipment before it
is installed in the MSEE proceseing fecility -

_ Demonstration of the continuous fluorinator and the partial decon-
tamination of fuel sslt from the MSRE in & practicel vacuum still.are
required before FY-19T71 in order to begin construction of the plent.
Demonstration of the protactinium removal process on & small scale by
that time is desireble end is planned, but it is not essential. Such =
\ process significantly'improves the performance of & molten-salt reactor
&c & breeder. It is not & decisive factor in making ean ‘MSBR competitive
with aavanced converter or fast breeder reactors. '

Maintenance bevelopment11

The methods for maintaining:much of the radioactive equipment in
the MSBE will be similar to those used in the MSRE. This eliminates the
expensive ‘consideration and investigation of seversl alternatives, but
considerable development of tools, jigs, end fixtures will be necessary
because their design is closely releted to the design of the reactor
equipment. Several technigues new to the molten-selt reactor technology
are proposed to be investigated &nd some will be developed. One is remote
machining and welding of the main salt piping. A second is the remote
replacement of the graphite structure core. A third is remote machining
and welding of seal welds or closure welds on the cover of the reactor
vessel and on the plenums. A fourth is the remote replacement of the
primary heat exchanger and possibly the plugging of heat exchanger tubes
in place or in & hot cell, depending on the design of the exchanger. The
welding end brazing development is & Joint Materlals Development end.
Maintenance Development effort. The program is plenned to demonstrate
by the end of FY-1970 the feasibility of making the essentiel joints in
the reactor system by remote brazing or welding or‘by other>methods pro-

. posed by the designers. .

| Physics Progrem?

Because the molten-salt breeder reactors are thernol.reectors,‘make
use of circulating fuels that are easily adjusted in fissile concentration,
and are of simple configurations, they do not require an eleborate physics
 

C

a4

81

program, Some work is needed to obtain better cross-section data. Studies
are required of the dynamics characteristics of the reactors and methods
of flattening the power distribution and some development of codes will be
necessary. Physics experiments will consist primarily of a few lattice
substitution measurements in the High-Tempereture Lattice Test Reactor

and the Physics Constants Test Reactor at the Pacific Northwest Laboratory.
' The program is planned to resolve by FY-19T71 all physics questions con-

cerning the performance of molten-salt reactors &s breeders. Work after

~that time will be mostly concerned with refining the physics calculations

and preparing for: the physics experiments associated with startup of the

MSBE.

Safety Program?s
The studies of safety of molten-salt reactors have, 4in the past, been

‘limited to the safety analysis of the MSRE. A thorough analysis is re-

quired of the safety problems of the large breeder reactors, primarily in
describing potential accidents, their consequences, and methods of pre-
vention.. Experimental investigation of specific problems such as release

of fission products from salt under accident conditions. and release of
pressure produced by discharge of supercritical steam into the intermediate
- coolant system will be made when the conditions are properly esteblished

by the enalysis. The analytical work end essential experiments can be
completed easlly as the reactor is designed. No problems are presently
foreseen that would lead to serious questioning of the feasibility of
properly containing and safely operating molten-salt reactor plants.

 
 

 

1.

2.
' 0RNL-3686 (Jenuary 1965).

13.

82

{References S ' ' \aj

Report of the Fluid Fuel Reactors Pagk Force , US=-AEC Report 'I’ID-—BSO’T,
(February 1959).

M. W. Rosenthal et al., A Com_parative Evaluation of. Advanced Converters y

Paul R. Kasten, E. S.. Bettis, Roy C. Robertson, Desi Studies of 1000-
Mw(e) Molten-Salt Breeder Rezctors, ORNL-3996 (August 1966).

M. W, Rosenthal et al., A Comparative Evaluation of Advanced Converters ,

ORNL-3686 (January 1965).

C. D. Scott end W. L. Carter, Preliminary Design Study of a Continuous
Fluorinetion-Vacuum Distlllation flstem for Regenerating Fuel and
Fertile Streams in a Molten Salt Breeder Reactor, 0RNL-3791 January
1966). ._ T

Dunlep Scott and A. G. Grindell, C onents and Systems Development
for Molten-Salt Breeder Reactors g ORNL-'.'EM-]BB'S iJune 30, 1%7;."

J. R. Tallackson, R. L. Moore, S. J. Ditto, Instr\mentation and

Controls Development for Molten-Salt Breeder Reactors, ORNL-13§6
(May 22, 196T).

H. E. McCoy and J. R. Veir, Materials Development for Molten?-Salt
Breeder Reactors, ORNL-TM-185E (June 1967).

#y

W. R. Grimes, Chemical Research and Development for Molten-Salt
Breeder Reactors ORNL-TM-15853 (June 1967)

W. L. Carter and M. E. Whatley, Fuel and Blanket Processing Develop-
nment for Molten-Salt Breeder Reactors, ORNL-TM-1852 lJune .1%‘75
Ro'bert Blum'berg s Maintenance Development for Molten-Salt Breeder

Reactors, ORNL-TM-1859 (June 30, 1967).

A. M. Perry, %fl/sics Program for Molten-Salt Breeder Reactors,
J _

ORNL-TM-1857 (June 1967).

Paul R. Kasten, Safety Pro for Molten-Salt Breeder Reactors,
ORNL-'I'M-lBSB (June '91', 1%4( im";n“"“w““"—_‘— - -
 

*
/

wi

Wk

1-50.

- 51.
52.
23

5k,
55 .
56.
5T,

58‘
59.
60.
- 61.
62.
63.
6,"‘0
65.
66.
67.
68.
69.

700

Tlo

T3«

Th.
750
760
7T,

78.

9.
81,

8o

- 91.

- 92,

- 93.

ok, S
-H. A, Friedman

%.
97.

J. M. Chandler - |

~J. R. Engel
E. P. Epler

L. M. Ferris  °

83
Internal Distribution

MSRP Director's Office
Room 325, 920L4-1
R. K. Adams

- G. M. Adsmson

R. G. Affel
L. G. Alexander

‘R. F. Apple

Cc- F. Baes

J. M. Baker .

S. J. Ball

W. P. Barthold
H. F. Bauman

S. E. Beall

M. Bender

E. S. Bettis

F. F. Blankenship’
R. E. Blanco

J. 0. Blomeke

R. Blumberg

E. G. Bohlmann

C. J. Borkowski - -
G. E. Boyd '
J. Braunstein

M. A, Bredig

‘R. B, Briggs

H. R. Bronstein
G’o D . Br\mton
D. A. Canonico
S. Cantor

W. L. Carter

G. I. Cathers

E. L, Compere

‘W. H. Cook

" L. T. Corbin ~

- J. L, Crovley .
F. L. Culler

_ - Je. M. Dale

B - D. G. Davis

88,
89,
90.

S. J. Ditto
A. S. Dworkin

D. E. Ferguson
A. P. Freas

J. H. Frye, Jr.
C. H. Gabbard

98.

99.

100.

101.

102,

103.
10k.
105.
106.

107.

108.
109.
110.
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113.

114,
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119.
120.
121.
122,
123,
124,

:125.
126,
127.

128.

129,

130.
131.
132.

- 133.
3.

135 .

136,
_‘ 1380 '

- 139,
1ko,

k1,
142,

1k3.

14k,
1k45.
1k6,

“B.
| E.

R.

“G.

H.
A,

H.

G.
S-.

'No

A.
G.
R.

-W.

W.
L.

Gallaher
Goeller
Grimes
Grindell
Guymon
Hannaford
Harley
Harmon
Harrill
Haubenreich
Heddleson
Herndon
Hightover
Hof'fman
Horton
Hudson

Inouye

- H.

R.

d.

T.
J.

R.

W.

Jordan
Kasten
Kedl
Kelley
Kelly
Kennedy
Kerlin
Kerr.

‘Kirslis

Krakoviak

. Krewson

Lamb
Lane
Lindauer
Litman

Lundin

Lyon

- MacPherson
, MacPherson
‘Martin
‘Mathews
Matthews

McClung

‘McCoy.

McDuffie
McGlothlan
McHargue

McNeese

Meyer
Moore
Nichols
Nicholson

 
 

147. L. C. Oakes 170. I. Spiewsk

148. P. Patriarca 171. R. C. Steffy

149, A. M. Perry 172. H. H. Stone

150. H. B. Piper 173. J. R. Tallackson

151. B. E. Prince 174. E. H. Taylor

152. J. L. Redford 175. R. E. Thoma

153. M. Richardson 176. J. S. Watson

154, R. C. Robertson - 177. C. F. Weaver

155. H. C. Roller 178. B. H. Webster

156. H. C. Savage 179. A. M. Weinberg

157. C. E. Schilling 180. J. R. Weir -

158. Dunlep Scott 181. W. J. Verner

159. H. E. Seagren 182, K. W. West

160. - W. F. Schaffer 183. M. E. Whatley

161. J. H. Shaffer 184, J. C. White

162. M., J. Skinner 185. L. V. Wilson .

163. G. M. Slaughte " 186, G. Young -

164. A. N, Smith - 187. H. C. Young =~ )
165. F. J. Smith 188-189. Central Research Library
166. G. P. Smith - 190-191. Document Reference

167. 0. L. Smith Section, Y-12

168. P. G. Smith 192-201. Laboratory Records

169, W. F. Spencer 202. Laboratory Records (RC)

1

External Distribution

203-204. D. F. Cope, Atomic Energy Commission, RDT Site Office
205. A. Glambusso, Atomic Energy Commission, Washington
206. W. J. Larkin, Atomic Energy Commission, ORO

207-221. T. W. McIntosh, Atomic Energy Commission, Washington
222, H. M, Roth, Atomic Energy Commission, ORO .

223-224. M. Shaw, Atomic Energy Commission, Washington
225, W. L. Smalley, Atomic Energy Commission, ORO.

226. R. F. Sweek, Atomic Energy Commission, Washington

227-241. Division of Technicel Information Extension (DTIE)

- 242, Research and Development Division, ORO
243-24Y4, Reactor Division, ORO