LEGAL NOTICE 1 I 

This report was prepared as an account of work sponsored by the United 

States Government. Neither the United States nor the United States Atomic 

Energy Commission, nor any of their employees, nor any of their 

contractors, subcontractors, or their employees. makes any warranty, 

express or implied, -or assumes any kgal liability or responsibility for the 

accuracy, completeness or usefulness of any information. apparatus, product 

or proms disclosed, or represents that its use would not infringe privately 

owned rights. 
AN EVALUATION 

OF 

THE MOLTEN SALT BREEDER REACTOR 

[----N 0 1 IC E 

This report was prepared-as an account of work 

sponsored by the United States Government. Neither 

the United States nor the United States Atomic Energy 

Commission, nor any of their employees, nor my of 

their contractors, subcontractors. or their employees, makes any warranty, express or implied, or assumes any 

legal liability or responsibility for the acc~~acy, com- pleteness or usefulness of my information, apparatus 

product or process disclosed, or represents that its US; 

would not infringe privatelv owned richts 

I I 

WASH- 12 2 2 

UC-80 . 

FO~ de by the Superintendent of Documents, U.8 0 veroment Prlntfng Ofaee 

Washfngton, D.C., 20402 *; $8 

v4 I)ISTRIBUTION OF THIS DOCUMENT IS UNLlMn 
This report was prepared as input to the Office 

of Science and Technology's Energy Research and 

Development Study conducted through the Federal 

Council for Science and Technology. 

represent the views of the panel members and not 

necessarily those of the Office of Science and 

The contents 

I Technology. 
TULE OF CONTENTS 

Page 

I. INTRODUCTION .0000..00..0o.o000..000000o0000000ooo0oo*ooo0ooooo 1 \ 

I I. SUMMARY ........ ............................................. 8 

111 . RESOURCE UTILIZATI .......................... 10 

IVo HISTORICAL DEVELOPMENT OF MOLTEN SALT REACTORS ................ 12 

V. MOLTEN SALT BREEDER REACTOR CONCEPT DESCRIPTION ............... 14 

VI . STATUS OF MSBR TECHNOLOW ..................................... 19 

A . 

B . MSRE - The Reference Point €or Current Technolwv ........ Continuous Fuel Processing - The Key to Breeding ......... 19 

22 

1 . ic pment ....................... 22 . 

2 . Fuel P uta1 Materials ............... 25 

C . Molten Salt R ...... 26 

1 . Fuel and Coolant Salts ............................. 27 

2 . Reactor Fuel Containment Materials ................. 30 

3 . Graphite ........................................... 33 

4 . Other Structural Materials ......................... 34 

D . Tritium . A Problem of Control ........................... 35 

E . Reactor Equipment and Sktems Development ................ 38 . 1 . Components ......................................... 38 

2 . Svstems ............................................ F . Maintenance . A Difficult Problem for the HSBR ........... C . Safetv . Different es for the MSBR .................. H . Codes. Standard High Temperature DesiRn Methods .... IAL PARTICIPATI MSBR PRWvRAM 0 . . . CONCLUSIONS ............. ................................. REFERENCES ............. .................................. APPENDIX A ............. .................................. 

c 

41 

43 

45 

47 

48 

51 

53 

A- 1 

. 
LIST OF ThLES AND'FIGURES 

Tables 

I Selected Conceptual Design Data for a Large MSBR. 

I1 Important Dates and Statistics for the MSRE 

Page 

17 

111 Camparison of Selected Parameters for the MSRE and . 

1000 W(e) MSBB 21 

2. Flowsheet for Processing a Single-Fluid MSBR. 24 

-. 

.. . ~ .. , 

, 

.. . . .. . .. . 
AN EVALUATION OF TEE MOLTER SALT BREEDEREACTOR 

I. INTRODUCTION 

The Division of Reactor Development and Technology, USAEC, was assigned 

-.- .~.. 

the responsibility of~assessing the status of the technology of the. -( 

Molten Salt Breeder Reactor (MSBR) & part of the Federal Council of 

_ . : 

Science and Technology Research and Development Coals Study. In. 

conducting this review, the attractive features and problem areas 

~. _ 

associated wi-th the concept have been examined; but more impor-tantly, 

_ 

the assessment has beendirected~.to provide a view of the Atechnologp 

-. __I-- 

and engineering development efforts and the associated government and 

-;- ~. i 

industrial commitments which tiould be required to dkvelop the MSBR 

._. 

into-a safe, reliable and economic power source for central station 

applicatiou. 

The MSBR 'concept, currently '. &er study at the Oak Ridge National 

Laboratory (ORNL), is based on use of~'a.circulati& fluid f&l 

reactor coupled with on-line continuous‘ fuel processing. As presently 

envisioned, it would operate as a thermal spectrum reactor system 

~utiliti& a thorium-u&i& fuel cycle. Thus, the concept kuld offer 

the @otent&i; for bro&e&d u&izat$o+of.the nation's natural 

. 

~resources through operation of a b.reeder system employing &&her 

_ _~ _ 

prg~i~“‘-f;bN 0~ THIS DOCUMENT IS UIWMiTED 
-2- I 

sustaining, competitive industrial capability €or producing economical 

power in a reliable and safe manner. 

this objective is to gain public acceptance of a new form of power 

production. Success in such an endeavor is required to permit the 

utilities and others to consider the concept as a viable option for 

generating electrical power in the future and to consider making the 

LJ 

A basic'part of achievement of 

, 

heavy, long-term commitments of resources in funds, facilities and 

personnel needed to provide the transition from the early experimental 

facilities and demonstration plants to full scale commercial reactor 

power plant systems. 

Consistent with the policy established for all power reactor development 

programs, the NSBR would require the successful accompl$shment of three 

basic research and development phases: 

. An initial research and development phase in which the basic 

technical aspects of the MSBR concept are confirmed, involving 

exploratory development, laboratory experiment, and conceptual 

in which the engineering and man 

capabilities are developed. This includes the conduct of 

in-depth engineering and prooftesting of first-of-a-kind 

components, equipment and systems. These would then be 

incorporated into experimental installations and supporting 
-3- , 

LJ test facilities to assure adequate xtnderst&ding of design 

and performance characteristics, as well as to gain overall 

experience associated with major operational, economic and 

environmental parameters. As these research and development 

efforts progress, the technohgical uncertainties would:need 

to be resolved and decision points reached that would permit 

development to proceed with necessary confidence. When the 

technology is sufficiently developed and confidence in the 

-system was attained, the next stage would be the construc- 

.tion of large demonstration -plants. ;; 

. A.third ~phase in.which the utilities make large scale commitments 

to electric-generating plants by developing the capability to 

manage the design, construction, .test and operation of these 

power plants in a safe, reliable, economic, and environmentally 

acceptable manner. f < 

,' -, Significant experience with the Light Water Reactor- (LWR), the high 

Temperature Gas-cooled Reactor ~(HTGR) and the Liquid Metal-cooled Fast 

. 

Breeder Reac.tor (L?JFBR) has been gbined over the.past two~decades 

pertaining to the 'efforts that are~required to-develop and advances ~-. 

nuclear reactors to the point of public and commercial acceptance. 

This experience has clearly demonstrated that the phases of develop- 

_ - 

m&-and d&nonstration'should be simiiar'regardless of the-energy 

~ ~_- _.. _-. -~~ 

concept being exploredi 

~_ _ .~ . 

that the logicai.progression through each of 

a. 
-4- 

t.e phases is essential; and that completing the work through the 

three phases is an extremely difficult, time cunsuminp: and costly 

undertaking, requiring the highest level of technical management, 

professional competence and organizational skills. 

been demonstrated by the recent experience in the expanding LWR 

design, corm truction and licensing activities which emphasize clearly 

the need for even stronger technology and engineering efforts than 

were initially provided, although ihese kre 

cases for the *first. experiments and demonstration plants. 

This has again 

. 

The LWBR 

program, which is relatively well advanced in its development, tracks 

closely this LWR experience and has further reinforced this need a. it 

applies to the technology, development and engineerinff application 

areas . 

It should also be kept in mind that the large backlog of commitments 

and the shortage of qualified engineering and technical management 

personnel and proof test facilities 'in the government , in -2ndusttv and 

in the utilities make it even more necessary that all the reactor 

systems be thoroughly designed and tested before additional significant 

commitment to, and construction of, commercial power plants are . 

initiated. 

~ 

With regard to the XSBR, prelfmhary reactor designs were 

WASH-1097 ("The Use of Thorid In Nuclear Pawer Reactors") based upon .., - ~ 
-5- 

t 

JJ the information supplied by &IL. Two reactor design concepts-were 

- . :. 1..- - --- 

considered -Y a two fluid reactor In which the fissile and fertile 

,_ =. _-: __ .- i. _ ~.~ _. 

salts were. separated by graphite and a single fluid concept in which r- L .-- .' : ~-_<- i. 

the fissile and fertile salts 'were completely mixed. This evaluation i = 

Identified problem-areas requiring resolution through -conduct :of an .i _. 

intensive research and developm-ent program. since the publication of I .I 1: -. = ._- _- 

WASH-1097, all efforts related to the two fluid system have-been 

discontinued because of mechanical design problems and the development * * : ., j 

of processes which would, if developed into engineering systems, 

permit the,on-line reprocessing of fuel from single ,f_luid reactors. 

-. . _ 

.~ 

At present, the.*BR concept is essentially in the initial research ,; _ I .-=-_ ._ _ -._- _ -- 7~'~. - 

and development phase, with emphasis ,on the development of basic MSBR Z~ 

_. 

technology. The technology program is centered at ORNL where fair. 7. 

essentially all research and development on molten salt reactors has 1 --~ .z -1 ~. _ .- 

_ been performed to date. The program is currently funded at a level 

_- ', _ .; 

. 

of $5 million per year. ~Expenditures to date on molten salt reactor 

_ . .- -_.~ ; : -_ 

technology both for military and civilian power applications have 

-. ~, L 

amounted to approximately $150 million of which approximately $70 million 

_~. -- : _i _. ;.- -e 

has been in support of central station ptier plants. These efforts date 7 - .~ _ in ~. ..; .-- .-. . I ~~ . -. -~. I. _~.: .~. 

back to the 1940's. _ 

; _. -.~~ pi .;~~ ~_ _ 

In considering the MSBR for central station power plant application, it ..I~:. ~:- = ~~. pi : .- .- 7~- :~~ 

,- is noted that this concept has several unique and desirable features: ‘. _ * -- j - --~ 

at-the -same time, it is characterized b-y both complex technological and : _ in-__ ~I. - * ,r I .- ; .~ 
-6- 

\J' 

practical engineering problems which are specific t 

reactors and for which solutions have not been developed. 

this concept introduced major concerns that are different in kind and 

magnitude from those cammanly associated with solid fuel breeder 

reactors. 

further consideration of this concept for use as a commercial power 

Thus, 

The development of satisfactory experimental units and 

plant will require resolution of these as well 88 other problems which 

are common to all reactor concepts. 

c 

As part of the AEC's Systems Analysis Task Force (AEC report 

WASH-1098) and the "Cost-Benefit Analysis of the U.S. Breeder 

Reactor Program" (AEC reports WASH-1126 and WASH-1184). atudies 

were conducted on the cost and benefit of developing another 

breeder system, "parallel" to the LMFBR. 

sion reached in these studies is that sufficient information is 

available to indicate that the projected benefits from the LMFBR 

program can support a parallel breeder program. Hawever, these 

A 

The consistent concluresults are highly sensitive to the assumptions on plant capital 

costs with the recognition, even among concepts in which ample 

experience exists, that capital costs and especially small estimated 

differences in costs are highly speculative for plants to he built 

15 or 20 years from now. 

analyses based upon such c 

making decisions relative to the desirability of a parallel breeder 

is ques t ionab le 

ould constitute a major basis for 7 

effort. 
-7- -* 

-_ 

Lid Experience in reactor development programs In this country and abroad 

has demonstrated -that different organ;itations; in evaluating the pro.jected 

costs of introducing a reactor development-program and 'carrying it forward 

to the point of large scale commercial utilization, would arrive at. 

- different estimates of the methods,- scope of development and engineering 

efforts, and the costs and time required to’bring &at pro&m ko’a stage 

of successful large scale application&d publfc"acceptance. 

Based uponthe ARC's experience with other complex reactor development 

programs, it is estimated that a total government 1nvestmektu~‘to 

about 2 billion dollars in undiscounted direct costs* could be required 

to bring the molten salt breeder or any parallel breeder to fruitfon as 

a viable; commercial power reactor. A magnitude of funding up to .&Is 

level could be needed .to estab&h'the necessa&' technology and, 

._ it_ I 

~._. . . 

. 

engineering bases ; obtafn the required industrial capability; and 

_ 

advance through a series of test facilities, reactor experiments, and 

: _-~, _ 

demonstratfon plants to.a commercial MSBR safe and suitable to aeme 

, _ 

as a major energy option for central station power generation in the 1 

~. 

utility environment. 

_ _I -.. .- ~ 

~,.~ - 

w _. = .- . 

i 

'WASH-ilEi4- 

,- 

-.Updated (1970) Cost-Benefit Analysis of the U.S. 

Breeder Reactor Program, January 3972W 
-8- 

11. SUMMARY 

The MSBR concept is a thermal spectrum, fluid fuel reactor which 

operates on the thorium-uranium fuel cycle and when coupled with 

on-line fuel processing has the potential for breeding at a 

meaningful level. The marked differences in the concept as compared. ~ I 

to solid fueled reactors, make the MSBR a distinctive alternate. 

Although the concept has attractive features, there are a number of 

difficult development problems that must be resolved: many of these 

are unique to the MSBR while-others are pertinent to any complex 

reactor, sps tem 

The technical effort accomplished since the publication of WASH-1097 

and WASH-1098 has identified and further defined the problem areas; 

however, this work has not advanced the program beyond the initial 

phase of research and development. 

in several areas -(e.g., reprocessing and improved graphite), new 

problems not addressed in WASH-1097 have arisen which could affect 

the practicality of designing and operating a MSBR. 

major uncertainties relate to materials of construction, methods for 

control of tritium, and the design of components and systems along 

' 

Although progress has been made 

Examples a€ 

with their spec;ial handling, inspection and maintenance equipment. 

Considerable research and development efforts are required in order 

to obtain the data necessary to resolve the uncertainties. 

. 
-9- 

Assuming that practical soluti o these problems can be found, a 

further assessment would have to be made as to the advisability of 

proceeding to the next stage of the development program. 'In advancing 

to the next phase, it would be necessary to develop a greatly expanded 

industrial and utility participation and commitment along with tl 

substantial increase in'govermnent support. Such broddened involvement would require an evaluation of the %BR in terns of already 

existing commitments to other nuclear power and'high prioritp energv 

development efforts. 
-10- 

111. RESOURCE UTILIZATION 

It has long been recognized that the importance of nuclear fuels for 

power production depends initially on the utilization of the naturally 

occurring fissile U-235; but it is the more qbundant fertile materials, 

U-238 and Th-232, which will be the major source of nuclear puwer 

generated in the future. 

plutonium produced from U-238 offer the potential for high breeding 

gaiqs in fast reactors, and the potential to expand greatly the 

The basic physics characteristics of fissile 

utilization of uranium resources by &king feasible the utilization of 

additional vast quantities of otherwise uneconomic low grade ore. In 

a similar manner, the basic-physics characteristics of the thorium 

cycle will permit full utilization of the nation's thorium resources 

while at the same time offering the potential for breeding in thermal 

reactors . 

The estimated thorium resemes are sufficient to supply the world's 

electric energy needs for many hundreds of years if the thorium is 

used in a high gain breeder reactor. It ie projected that if this 

quantity of thorium were used in a breeder reactor, approximately 

1000 Q (1 Q = 10l8 Btu) would be realized from this fertile material. 

It is estimated that the uranium reserves would also supply 1000 Q* 

of energy if the uranium were used in LMFBRs. In contrast, only 20 Q 

*Uranium recoverable at U308 price up to $lOO/lb. 
-11- 

would be available if thor 

an advanced converter reactor because the reactor would be dependent 

e fertile material in 

upon U-235 availability for fiss inventory make-up (Note : 

conservative estimate is that between 20 and 30 Q will be 

electric power generatio nw and the year 2100.) 

.- 

\ 
-12- 

IV. HISTORICAL DEVELOPMENT OF MOLTEN SALT REACTORS 

The investigation of molten salt reactors began in the late 1940's as 

part of the U.S. Aircraft Nuclear Propulsiod (ANP) Program, 

the Aircraft Reactor Experiment (ARE) was built at Oak Ridge and in 1954 

it was operated successfully for nine davs at power level@ up to 

2.5 MW(th) and fuel outlet temperatures up to 1580.F. 

mfxtute of NaF, ZrP4, and UF4. 

Subsequently, 

The ARE fuel was a 

The moderator was BeO.and the pipinn and 

vessel were constructed of Inconel. 

In 1956, ORNL began to study molten salt reactors for application as 

central station converters and breeders. These studies concluded thatl 

graphite moderated, thennal spectrum reactors operating on a thoriumuranium cycle were most attractive for economic power production, Based 

on the technology at that time, it was thought that a two-fluid reactor 

in which the fertile and fissile salts were kept separate was required 

in order to have a breeder system. 

a breeder, appeared simpler in design and also seemed to have the 

The single fluid reactor, while not 

potential for low power costs. 

Over the next fw years, ORNL continued to studp both the two fluid and 

single fluid concepts, and in 1960 the' design of the single fluid 

'8 MW(th) Holten Salt Reactor Experiment (MSRE) was begun. The MSRE was 

completed in -1965 and operated successfully during the period 1965 to 

1969. The %RE experience is treated in more detail in a later section. 

U 
Concurrent wlth.the construction of the MSRE;OR&performed research 

. 

and development on means for proceseinp; molten salt fuels. In-1967 

new discoveries were made which suggested that a single fluid reactor. 

could be combined with continuous on-line fuel processing; to become a 

. . I. 

breeder sys tern. Because of the mechkcal design prob1en-m of the two 

fluid concept and the laboratory-&ale development of process& whichwould permit20n~lke reprocessing, it was deierdninkd that 8 shift in 

emphasis to the single fluih breeder concepk should be make; this 

e_._- .; 

system fs being studied at the present. -2 

/’ 
-14- 

V. litoLTEN SALT BREEDER REACTOR CONCEPT DESCRIPTION u 

The breeding reactions of the thorium cycle are: 

B ,233v B 233P€4 22 min. e- 27.4d 232~ + n-233~ 

Because of the number of neutrons produced per neutron absorbed and the 

mall fast fission bonus associated with U-233 and Th-232 in the 

thermal spectrum, a breeding ratio only slightly greater than unity is 

achievable. 

necessary to remove the bred Pa-233 and the various nuclear poisons 

In order to realize breeding with the thorium cvcle it is 

produced by the fission process from the high flux region as quickly as 

possible. 

of Pa-233 and the nuclear POiSOnS (e.g. Xe-135 and the rare earth 

The Molten Salt Breeder Reactor concept permits rapid removal 

elements). The reactor is a fluid fueled system containing uF4 and 

ThF4 dissolved in LiF - BeF2. 

graphite moderator where the nuclear reactions take place. 

The molten fuel salt flows through a 

A side 

stream is continuously processed to remove the Pa and rare earth 

elements, thereby permitting the achievement of a calculated breeding 

ratio of about 1.06. 

The MSBR is attractive because of the following: 

1. Use of a fluid fuel and on-site processing would eliminate the 

problems of solid fuel fabrication and the handling, and 
-15- 

LJ shipping and reprocessing of spent fuel elements which are 

associated with all other reactor types under active * 

consideration. I 

2. MSER operation on the thorium-uranium fuel cycle would help 

conserve uranium and thorium resources by utilizing thoriumreserves with high efficiency. ._ ; ~, 

, 

3. 

I 

The MSBR is,, projected to have attractive fuel cycle costs. 

j The majors uncert,ainty in the fuel cycle cost is associated 

with the continuous fuel processing plant whi.ch has not been I. 

developed. 

. . .- 

I 

4. The safety issues ass&a&d with the MSBR are genetallv 

different .from those of &id -fuel reactors. Thus, there 

might be safety advantages for the KSBR when considering 

major accidents. An accurate- assessment of MSBR safety is 

not possible today because of the early state of development. 

5. Like other advanced reactor systems such as the LMFBR and \ 

- HTGR, the MSBR would employ modem'steam technologv for power 

generation with high thermal efficiencies. This would reduce 

the amount of waste heat to be discharged to the environment. 
-16- 

Lid Selected conceptual design data for a large MSBR, based primarily on 

design studies performed at ORNL, are given in Table 1. 

There are, however, problem areas associated with the MSBR which wst 

be overcome before the potential of the concept could be attaiked* 

These include development of continuous fuel processing, reactor and 

processing structural materials, tritium control methods, reactor 

equipment and systems, maintenance techniques, safety technolonv, and 

MSBR codes and standards. Each of these problem areas will now be 

evaluated in some detail, 

which was demonstrated by the Molten Salt Reactot Experiment (“2) 

during its design, construction and operation at Oak Ridge and the 

conceptual design parameters presented In Table I and in Appendix A. 

A conceptual flawsheet for this system is shown in FiRure 1. 
-17- 

Table I . 

Selected Conceptual Design Data for a Large KSBR 

Power, W(e) 1000 

1 Pawer, W(th) 2240 

Steam System 3500 psia, 100O0F, 

44% net efficiencv 

7 * 72% LIF, 16% BeF2, ~ 

lng and Vessel 

%dera tor Sealed Unclad Craphlke 

Breeding Ratio 

Specific Fissile Fuel 

ded Doubling The, Yea -- 

Core Temperatures, OF 1050 Inlet, 1300 outlet 
\ 

\ 
-19- 

VI. STATUS OF:MSBR .TECUNOLOGY 

A. MSRE- The Reference Point for Current,Technology 

I~ .~. 

The Zlolten Salt Reactor Experiment (MSRE) was begun in .1960 at 

ORNL as part of the Civilian Nuclear Power Program. Th? purpose 

of the experiment was to demonstrate the basic feasibility of 

molten salt power'reactors. All objectives of the experiment 

were achieved-during its successful operation'from June 1965 to 

December 1969.. These included the distinction of becoming the 

first reactor in the world to operate".solely on U-233. Some of 

the more signifikt dates and statistics pertinent tom the MSRE 

are givenin Table II. 

In spite of thesuccess of the MSRE, there are -many areas of molten 

salt technology which must be expanded and developed in order to 

proceed fram this small non-breeding experiment to a safe, reliable, 

and economic 1000 W(e) MSBR with-a 30-year life. To illustrate 

,. -. 

. 

this point, some of the most important differences in basic design 

. 

and performance characteristics between the MSRE and a conceptual 

lOOO.hW(e) MSBR are given inTable III. Scale-up would logically 

be accomplished through development of reactor plants of increasing 

size. Examination of Table III provides an appreciation of the 

scale-up requirements in ,going from the MSRE to a large MSBR. Some 

problems associated with progressing from a smaJ1 experiment to a 

commercial, high performance power plant are not adequately 

i 
-20- 

Table 11. 

Dates: , 

i 

Design initiated . . . . . . . . . . . . . . July 1960 

Critical with 235U Fuel . . . . . . . . . June 1, 1965 

operation at full power - 8 MtJ(th) . . May 23, 1966 

Complete 6-month run . . . . . . . . . . . . March 20, 1968 

End Operation with 235U fuel . . . . . . . March 26, 1968 

Critical with 233U fuel . . . . . . . . . October 2, 1968 ~ 

Operation at full power with 233U fuel .. . . January 28, 1969 

Reactor operation terminated . . . . . . . . December 12, 1969 

Statistics : 

Hours critical . . . . . . . . . . . . . 17,655 

Fuel loop time circulating salt (Hrs). . . 21,788 

Equiv. full power hours with 235U fuel . . . 9,005 

Equiv. full power hours with 233U fuel . . . 4,167 

LJ 
-21- 

Ld Table III. 

11 Comparison of Selected Parameter8 of the MSRE and 1000 MN(e) MSBR - > 

General 

MSRE- MSBR Y 

v ; 

ThermalPcwer,MW(th) s 2250 ' 

Electric Pwer, MN(e) 0 1000 

Plant lifetime, year8 /: .-, ~ 4- ~_ ~30 ~_ 

Fuel Proceasing Scheme Off-line, batch -- On-line, continuous 

: -7 I .: processing:; -I processing - 

. . . Breeding Ratio ~:_ IT k88 thSUl 1.0 1,06- 

~(No Th present) .- 

Reactor 

Fuel Salt 7 7 

.-. 

_ LiF-BeFZTZrF4-UF4 

= Lif-BeF2TThF4-UP4 

Moderator '. ~-Unclad, Unclad, 

unsealed 'graphite- sealed graphite 

Reactor Vessel Mater;al Standard Hastelloy-N Modified Hastelloy-N 

Pwer DeAity,~RW/iiter 2.7 -- 22 

Exit T&perature, l F -~- -~.'- ~. 

- 

1210 

;=c, ; 

1300- 

Temperature Rise ACrO88 Core.1OF 40~.. - _ ~m.250 

Reactor Vessel Height, Ft. 0 -20 

Reactor Vessel Diameter, Ft. 5 22 

Vessel Design Pressure,-psia'~. 65 ~--' - 75.- -. 'I* 

,8.3-x lO14 A 

Neutrons/m‘+sec 

Other Components and Systems Data m.mz,I-=- _ = _ . ; -*-_ - 

Number of Primary Circuits Jm. ' ~' 4 _ .:.. -,: .I. - 

~Fuel Salt Pump Fl&; j$m * ". y1200--. r. I L,. .-16,0& 

Fuel s&t mp @ad,. ft,::; I’ ~48;s ~. _ : - ~. 150. ~~ ~: 

Secondary Coolant Salt 7 LIF-BeF2 NaF-NaBF4 

Numbers Numbers of Secondary Secondary ~Circufts ~Circufts :-l _ ~: ,._ 4. 

Secondary-Salt Pump Flaw, Epm -850 -_ 

,. _~ - :~ 1 

20,000 

_ 

Secondary~Salt Seqzmdary~Salt Pump Head, ft. 78 300 

Nuniber~of Number~of Steam Gener&ors ~0 i6 . 

Lb Steam Generator Capacity, MW(th) 0 121 

A/ Based on information from "Conceptual Design Study of a Single Fluid Molten 

Salt Breeder Reactor," ORNL-4541, June 1971. 
-22- 

w 

represented by the comparison presented in the Table. -Therefore, 

it is useful to examine additional facets r of MSBR technology in 

more detail. 

B. Continuous Fuel Processing - The Key to Breeding 

In order to achieve nuclear breeding in the single fluid NSBR it 

is necessary to have an on-line continuous fuel processing system. 

This would accomplish the following: 

. Isolate protactinium-233 from the reactor environment so it 

can decay into the fissile fuel isotope uranium-233 before 

being transmuted into other isotopes by neutron frradiation. 

. Remove undesirable neutron poisons from the fuel salt and 

thus improve the neutron economy and breeding~performance 

of the system. 

. Control the fuel chemistry and remwe excess uranium-233 

which is to be exported from the breeder system. 

1. Chemical Process Development 

The Oak Ridge National Laboratory has proposed a fuel 

processing scheme to accomplish breeding in the MSRR, and 

the flowsheet processes involve: 

a. Fluorination of the fuel salt to remove uranium as UF6. 
-23- 

b. Reductive extraction of protactinium by contacting the 

salt-with a mixture of lithium and bismuth. 

c. Metal transfer processing to preferentially remove the 

. 

rare earth~fission product poisons which would-otherwise 

hinder breeding performance. 

The fuel proc"eseing system shown in Fig. 2 is in an earlv stage 

. of development at present and this tvpe of system has not been 

demonstrated on an. operating reactor. By comparison, the MSRE 

required only off-line, batch fluorination to recover uranium 

from fuel salt.. I 

At this time, the basic chemistry involved in the MSBR 

processing scheme has been demonstrated in laboratory scale 

_ i 

experiments. Current efforts at O&k Ridge are being-directed 

toward development of subsystems incorporating many of the 

_- 

required processing steps. Ultimately a complete breeder 

processing~ experiment would be requfred to demonstrate the 

system with all~the chemical conditions and operational 

requirements which would be encountered with any MSBR. 

Not shown on the flowsheet is a separate processing svstem 

_ 

which would require injecting helium bubbles into the fuel 

~~ .~~ 

.- 
a 

P 

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W 

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W 

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I J 

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I 

I 

I bl!! 

W 

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0 

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a 

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I 

I 

I 

I 

I 

0 

1 I

a 0 -ti g --- 

W 

J 

3 

0 

P 

a 

e 

2 

W 

0 

3 

V 

c 

W 

W 

x 

c 

a 

a 

V 

c 

2 

a 

z 

e 

I 

W 

I 

W 

e 

U 

J 

? 

LL 

z 

0 

F 

2 

0 

a 

1 I 

a 

fn 
salt, allowing thgn to~~ciryzulate in the reactor system until 

~. they -collect flsslon~product xenon, and then removing the 

bubbles and xenon from the reactor system.. Xenan.ls a-highly 

undesirable neutron poison which will hamper breeding -perf&n- 

- ante by capturing neutr& which -would otherwise breed new 

~. -fuel. This concept for xenon stripping was -demonstrated~ln \ 

prlnclple.by~the~MSRE, althouih more efficient-and controllable : 

stripping syst+ns[ willmbe-desirable ior <the MSBR. The kenon 

polsonlng in the MSRE_was ~reduceh by a. factor of six by xenon 

stripping; the goal for the MSBR ls a factor of ten reduction. 

_ : 

2. Fuel Proceselng~Structural Materials : 

Aside fromthe chemical processes themselves,. there are also 

development requirements assoclated.wlth containment Fterlals 

for the fuel.processlng systems. -19 ~artlcular~~llquid bismuth 

presents dif flcult compatlbl<y p?oblems wlth most structural 

metals., and present effortsTare concentrated on using moly)denum 

and graphite for contalnlng,blsmuth. U>fortunately, both 

-molybdenum and-graphite -are;-difficult to use for +uch~englneerlng 

applications. T)us, it w&Xl-be necessary to~d&@op.improyed 

: ~ techniques-for fabrication-a@ joining-before their.use 1s. 

. ~~ 

~~ .posslble ln.the'reproi~filnlF; system._ -_ -, ~. ._ -'- 

__ .- j .-. .I -i. 
-26- 

A second material8 problem of the current fuel processing system 

ia the contafnment for the fluorination step in which uranium is 

volatilized from the fuel salt.' The fluorine and fluoride salt 

mixture f9 corrosive to most structural materials, including 

graphite, and present ORNL flowsheets show a "frozen wall" 

fluorinator which operates with a protective layer of frozen 

fuel salt covering a Aastelloy-N vessel wall. 

would require considerable engineering development before it is 

truly practical for use in on-line. full proceasing systems. 

This component 

C. Molten Salt Reactor Design - Materials Requirement8 

In concept, the molten salt reactor core is a comparatively 

uncomplicated type of heat source. The MSRE reactor core, for 

example, consisted of a prhmatic structure of unclad graphite 

- 

moderator through which fuel salt flowed to be heated by the 

self-sustaining chain reaction which took place as long as the 

salt was in the graphite. The entire reactor internals and fuel 

salt were contained in vessels and piping made bf Haetelloy-N, R 

high-strength nickel base alloy which was developed under the 

Aircraft Nuclear Propqlsion Program. Over the four-year lifetime 

of the ERE, the reactor 'structural bterials performed satisfactorirp for the purposes of the experiments although operation 

of the MSRE revealed possible problems with long term use of 
t 

-27- 

hi Hastellay-N in contact with fuel salts containing fission 

products. 

' 

The MSBR application is mare demanding in many respect8 than the 

MSRE, and additional d 

several areas of mater 

could become available. 

ork would be 'required in 

om before suitable materials 

- 

1. 

7 The ERE fuel salt was a mixture of 

proportions of 

LiF-BeF-ZrP4-UF4 in , 

.0-29.1-5.0-0.9 mole X, respectively. 7 

Zirconium fluoride included ae protection against U02 

precipitation should inadvertent oxide contamination of the s 

epatem occur. MSRE operation indicated that control of 

oxides was not a major problem and thus it is not considered 

necessary to include zirconium in future molten salt reactor 

fuels. It should also be noted that the MSRE fuel contained 

whereas the proposed MSBR fuels would include 

for breedfnk. With the 

possible except 

lsf actorily throughout the 

rrently proposed by ORNL, would be a 

in proportions of 71.7-16-12-0.3 
-28- 

Lid 

mole X, respectively. This salt has a melting point of about 

930.F and a vapor pressure of less then 0.1 mm Hg at the mean 

operating temperature of llSO'F. 

the density and 10 tbes the viscosity of water. 

conductivity and volumetric heat capacity are comparable to 

It also has about 3.3 times 

Its thermal 

water. 

- 

The high melting temperature is an obvious limitation for a 

8ystem using thfs salt, and the MSBR is limited to high 

temperature operation. 

must be enriched in Li-7 in order to allow nuclear breeding, 

since naturally occurring lithium contains about 7.5% Li-6 . Lid is undesirable in the MSBR because of its tendency to 

capture neutrons, thus penalizing breeding performance. 

fa addition, the lithium component 

- 

. 

The chemical and phyeical characteristics of the proposed 

MSBR fuel mixture have been and are being investigated, and 

they are reasonably well known for wirradiated salts. The 

major unknowns are associated with the reactor fuel after it 

has been irradiated. 

the behavior of fission products. 

fission product behavior ie important to plant safety, 

operation, and maintenance. While the MSRE provided much. 

useful iafonaation, there is still a nee 

For example, not enough is-known about 

The abil3ty to predict 
-29- 

$artGzularly with regard toi‘the- fate of the- so-called "noble 

-. ,. - metal" fission prqducts’such aS mol~bd&ium, niobiti and 

others which are cenerated-%n substantial quix%i~ies and 

~_ &ose behaVioriti tkisystem is not well ‘unders toad. 

_I -.- _ - .--‘I.. 

.-- A more &%plete'underst~anding of the uhy&cal/chenical 

--~-characteristics of the-irradiated fuel salt is also needed. 

As an illustrationof this point, anomalous powef pulses were 

~. 

observed- d&&g early‘operation oi the ?¶SRE with U-233 fuel 

: : \ : ' _*. ._ 

&hich were attributed to unusual behavior of helium pas 

..= 

' bubbles as they circulated through the reactor. This 

_,.- j -.- ~- 

behavior is bkeved to have been due to some physical and/or 

_- 

chemical characteristics of the fuel salt which were never 

fully understood. Out-of-reactor Vork on molten fuel salt 

f&ion product chemistry is currently under way. Eventually, _ -~.~~~ 

the behaviorSVof the f&l salt-would need to be confirmed in an .I "~7 

operating reactor. 

.-The cools& salt in the secondary secondary system of the WRE was of _ - 

. . ~. .: . . 

molar composition 66% 'L$F-34XBeF2. While this coolant 

/ 

performed satisfactorily (noldetectable.corrosidn or~reactfon 

,-could-be observed in-the secondary system), the salt-has a 

_ _ high melting. -temperature =(850?F) and la relatively expensive. 

Thtis, it may not be-the appropriate choice- for power reactors 

performed satisfactorily (noldetectable.corrosidn or~reactfon 

,-could-be observed In-the secondary system), the salt-has a 

_ _ high melting..temperature=(8500F) and la relatively expensive. 

Thtis, it may not be-the appropriate choice- for power reactors 
-30- 

for two reasons: (1) larger volumes of coolant salt will be LJ 

used to generate steam in the MSBR, and (2) salt temperatures 

in the steam generator should be law enough, if poesible, to 

utilize conventional steam system technology with feedwater 

j 

temperatures up to about 550'F. 

less affected by the coolant salt melting temperature since 

The operation of MSRE was 

it dumped the 8 MW(th) of heat via an air-cooled radiator, 

The high melting temperatures of potential coolant salts 

remain a problem. The current choice is a eutectic mixture 

of sodium fluoride and sodium fluoroborate with a molar 

composition of 8% NaF-92% NaBF this salt melts at 725.F. 

It is cmparativelv inexpensive and has satisfactorv heat 

transfer properties , 

4; 

However, the effects of heat exchanger leaka between the 

coolant and fuel salts, and between the coolant salt and 

steam systems, must be shown to be tolerable. The 

fluoroborate salt is currently being studied with respect 

to both its chemistry and compatibility with Hastelloy-N. 

2. Reactor Fuel Containment Materials , 

A prerequisite to success for the MSBR would be the clbilitv 

to assure reliable and safe containment and handling of molten 
-51- 

u fuel salts at ail: times dur'ink the life of the reactor. It 

' w&d be necessary,; therefore; to "develop suitable contain- 

-merit materials -for MSBR application before plants could be 

con&ructed;- . . . 

-. 

- r .--- I -5 I

A serious question concerning compatability of Eastelloy-N with 

the constituents of .irradiated fuel salt was raised by the postoperation examlnation:of the-MSRE-in 1971. Although the MSRE 

materials performed satisfactorily for that sqstem during its 

operation, subsequent examintition of metal which was exposed to 

MSRE fuel salt revealed that the alloy had experienced intergranular attack to’cleptha ~of about 0.007 inch. The attack was 

not obvious. tit%1 metal specim&s were tensile tested, at which 

time &a&-*opened up as‘the metal was strained.~ Further 

examination-revealed that several fission products, including 

tellurium, had-penetrated the metal to depths comparable to 

‘those of the cracks. At the present'time, it is thought that 

the intergrantilar-attack,was due to the.presence of tellurium. 

Subsequent laboratory tests have verified that tellurium can 

produce', under~certainconditlons, interffranular cracking in 

.' aast&rop N* . .~ '- *, + -, ._ 

., - .~ 

.~ Although the limited penetration of cracks presented no problems 

for the-MSRE, concern & exists,with respect to the chemical 
-32- 

compatabilitv of Hastellay-N and MSBR fuel salts when subjected LJ 

to the more stringent MSBR requirements of higher power density 

and 30-year life. If the observed intermanular attack was 

indeed 'due to fission product attack of the Hastelloy-N, then 

this material may not. be suitable for either the piping or the 

vessels which would be exposed to much higher fission product 

concentrations for longer periods of time. Efforts are under 

way to understand and explain the cracking problem, and to 

determine whether alternate reactor containment materials 

should be actively considered. 

I 

In addition to the intergranular corrosion problem, the standard 

Hastelloy-N used in the MSRE is not auitable for use in the MSBR 

because its mechanical properties deteriorate to an unacceptable 

level when subjected to the higher neutron doses which would 

occur in the higher power density, longer-life MSBR. 

is thought to be due mainly to impurities in the metal which are 

transmuted to helium when exposed to therrnal neutrons. 

is believed to cause a deterioration of mechanical properties by 

The problem 

The helium 

its presence at grain boundaries wlthin the alloy. It would be 

necessary to develop a modified Hastelloy-N with improved irradiation resistance for the MSBR, and some progress is beinff made in 

that direction. It appears at this time that small additions af 

certain elements, such as titanium, improve the irradiation. 
-33- 

.- 

performance of Hastelloy-N substantially. Development work on 

modified. alloys with imprwed,irradiationresietance is 1 

currently under way. 

3. &phi& 

Additional developmental effort ont& problems'& required to 

-produce graphites suitable for VSBR &piicatian. The-first 1s 

associated with irradiation damage to graphite structures which 

results from fast neutrons. Under high neutron doses, -of the 

order of 1022neut.rons)~2 , most graphites tend to-become 

dimen&nally~unstable and grossswelling of the material occurs. 

Based on testsof- small graphl&sam$es at ORNL, the best. 

commercially available graphites at this time-may be usable to 

about 3 x i0 22 neutrons/cm2 , before the core graphite would have ~l_ 

. 

to. be replaced. This corresponds to roughly a four-year graphite 

-~ . 

lifetime-for the ORNL~reference design. Wile this might be 

i . .- 

acceptable, there are still uncertainties about-the fabrication 

I L 

and performance of large graphite pieces, and additional work 

-, 

would be required before a four-year life could be assured at 

the higher MSBR power densities~how being considered.. In anv 0 --.. -1 ;a : 

event, -there would be an obvious~econamfc;.+ncentive~t~~ develop 

longer lived graphites for MSBR application s$nce:a--four-year .- 

.I 

_ life for graphite is- estitiated to represent a fuel &le cost 

' _"'_ 

w , 
-34- 

penalty of about 0.2 mills/kw-hr relative to a system with W 

thirty year graphite life. 

The second major problem associated with Rraphites for MSBR 

application is the development of a sealing techniaue which 

will keep xenon, an undesirable neutron poison, from diffusing 

into the core graphite where it can capture neutrons to the 

detriment of breeding performance. 

may not be necessary to achieve nuclear breeding in the MSBR, 

the use of sealed graphite would certainly enhance breeding 

performance. The economic Incentives or penalties of graphite 

sealing cannot be assessed until a suitable sealing process is 

While graphite sealing 

developed . 

Sealing methods which have been investigated to date include 

pyrolytic carbon coatiag and carbon impregnation. Thus far, 

hawever, no sealed graphite that has been tested remained 

sufficiently impermeable tb gas at MSBR design irradiation 

doses, and research and develo in this area 1s continuing. 

4. Other Structural Materials 

i 

In- addition to the ctural materials requirements for the ! 

L 

reactor and fuel processing systems proper, 

components and systems which have special materials reauirementa. Such components as the primary heat exchangers and 
-3L _ : 

steam generators~must function"while in contact with two 

different working fluids. .~~~ 

,- 

At the present time, Hastelloy-N is cormldered to be the most 

promising material for'use in all salt containment systtie, 

including the secondary piping and components. Research to date 

indiicates that sodium fluoroborate and HaStellOy-N are colnpat- 

~_ / 

ible as long as the water content of the fluoroborate is kept _ : 

low; otherwise, accelerated Corrosion can occur. Additional 

testing would be needed and is underway. 

. . 

Hastelloy-Nil'ha8 not been adequately evaluated for service under 

a range of steam condition8 and whether it will be a suitable 

material for use in steam generator8 is still not kncm. 

D. Tritium - A Problem of Control 

Because of the lithium present in fluoride fuel salts, the present WBR 

concept has the inherent problem of generating trithun, a radioactive 

isotope of hydrogen. Tritium is produced by the- foll&ing reactions: 

ti1 'Li (nia) %. 

(2) 'Li (n,aa) 'H. 

Due primarily to 'these interactians, tritlum would be-produced at a rate 

of about 2400 curies/day in a 1000 We MSBR. This compares with about 
- 30 - 

40 to 50 curiesldrtv for li.rht water, Raa-cooled, and faqt hrecrter 

reactors, in which tritium Is produced primarilv as a low vield fisston 

product. 

is Renetallv in the rmEe 3509 to 5800 curies/dav, due to neutron interactions with the deuterium nresent in heavy water. 

Tritium productinn in heavy water reactors of cormarable size 

-= 

To' further compound tile prnhlem tritim diffuses readtlv throuzh 

Hactellov-N at elevate> tempetatuteq. 

to prevent tritium Cram dfffrrstnE throtlRh the piplnp 

As a tesri ,-. 

conlprinmts of 

the VSRQ system (such heat exchancars) and everrtuallv reachin? the 

3te.m system where ft mip.ht he discharr.ed th the environwnt ttftintd 

water. 

-. . 

The problem of trCtitm control in the TBP fs hef.n$ studied In detsil at 

r)!?NL. The followins ate beinr considered as potentfa1 mathoas for 

tr€t ium control : 

1. Exchangfnp the trttim far an-? hvdronen nresent fn the secondat-. 

coolant, therebv retaininp the trttltim tn the sacnnriarv cnolnnt. 

2. Usinp, coatinEs an metal surfaces fn order to lnhibft trittltm 

diffusion. 
. . 

-37- 

. 

L : 3. OperatIny, the reActor with-the salt more axid*zfn?,~ therebv ' I' 

causing, the form&ion of-tritium fluoride which~.couldS>be 

removed fn the off-Eas svstems.~ 

. 

4; Usfnpa different secondant coolant,. erP;., *odium or helium, and 

pracesstnfi this coolant to~remove tritium; 

1 . . 

S.- Using--another intermediate loop between the fluorahorate and 

~steam to "Setter" tritium-mm 

6.- Using duplex tubing fn either the heat exchanger or steam ~- 

generator with a purge pas between the-walls: 

Of these notentisl solutions, the use of an additional intermediate loop 

between the secondary -and steam.systems IS considered the most effective 

method technically. but -it Twould also be extiensive due tti the adbitirrnal 

, 

equipment required and the loss of thermal efflcfencv.- ; T 

-. 

From an economfc viewpoint,- the mostdesYrable solutlnn is one which 

does not. significantlv .coinplictite -‘the svstem, :kuch RS exchazipe of 

-tritium -for.-hydrogen ~present =in--the seconda?+ coolant;-~~Thls techninue 

is being investigated a<part of-the ORNL~efforts on‘ trttium chemistrv. 

The tritium retention problem may be eased bv the low nertneahilftv of 

.TI:.e .~ _ I. I m. .I- 

._ '~ . -*I 

. 
Oxide coatings which occur 011 steam generator materials in contact with 

steam, and this is also being investigated at ORNL. 

E. Reactor Equipment and Sfstems Development 

While the XSBR would utilize some existing engineering technology from 

other reactor types, there are specific components and systems for which 

additional development work is required. 

into account the induced activity that those components would accumulate 

in the MSBR system, i.e., special handling and maintenance equipment wouSuch work would have to take 

also need to be developed. 

with a number of these, such as fuel processing components and systems, 

but additional discussion is appropriate. 

The previous discussion has already dealt 

1. C0mP onents 

As indicated in the Table 111, a number of components must be 

scaled up substantially from the %RE sizes before a large MSBR 

is possible. The development of these larger components along 

with their special handling and maintenance equipment is probably one of the most difficult and costly phases of HSBR I 

development, 

components would need to be developed in order for any reactor 

system to be a success. 

However, reliable, safe, and maintainable 

The MSBR pumps would likely be similar in basic design to those 

for the MSRE, namely, vertical shaft, overhung impeller pumps. 
-39- 

-Substantial experience has been-gained over the years in the 

design,-'fabrication and operation of smaller salt pumps, but 

-the size Gould have~to be increased substantially for MSBR 

application. The deve.lopment and proof testingof such units 

~along with their handling and maintainence equipment ,and test 

facilities are expected to be costly and time consuming. 

The intermediate heat exchangers for the MSBR must perform with 

a minimum of salt inventory in order to improve the breeding 

performance by lowering the fuel inventory-. Special surfaces to 

enhance heat transfer would help achieve this, and more studies 

would be in order. ~Basedsn previous experience with other reactor ' 

systems6 it is believed-that theseIunits would require a difficult development and proof testing effort; I 

The steam generator for MSBR applications is probably the most 

difficult large component to develop since it represents an 

item for Which there has'been almost.no experience to.'date., It 

.is believed that a difficult develo$ment~and~proof testing pro- 

' -gram would'beneeded to provide reliable and maintainable units. 

m--As discussed -previously,~the.~hij3h'melting temperatures- of 

:- candidate secondary koola&si such ;as' sodium fluoroborate, 

present problems -of matching with conventional iteain system 

~technology~ -At this timeimcentral station so&r plants utilize 

j, ) 
-40- 

feedwater temperatures only up to about 550°F. Therefore, 

coupling a conventional feedwater syetemto a secondary 

coolant which freezes at 72S'F presents obvious problems in 

design and control. 

cations to conventional steam system designs to help resolve 

the problenk. Because of these factors, a etudy related to 

the design of steam generators has been initiated at Foster 

Wheeler Corporation. 

It might be necessary to provide modifiControl rods and drlves for the MSBR would also need to be 

developed. The MSRE control rods were air cooled-and operated 

inside Hastellop-N thimbles which protruded dawn into the fuel 

salt. 

higherepower densities involved. 

The MSBR would require more efficient cooling due to the 

Presumably rods and drives 

would be needed which pemit the rods to contact'and be cooled 

by the fuel salt. 

The salt valves for large MSBR's represent another development 

problem, although the freeze valve concept: which was employed 

successfully in the MSRE could likely be scaled up in size and 

utilized for my MSBR applications. Mechanical throttling 

vdlvek would also be needed for the MSBR salt ~ystems, even 

though no throttling valve was used with the MSRE. Mechanical 

shutoff valves for salt systems, if required, would have to be 

developed . 
.:. Other components whichwould require considerable engineering 

devklcjpment and testing include-the helium bubble generators and 

.gas strippers which are- +roposed.for use in removing the fission 

i product xenon from&the ~fuel salt. 1 Research and development in 

this area is cukently &der way as part of the technology 

." program at ORNLm 

" . : - .~ 

_- 

2. Systems i 

.The integration of all-‘required’components into a complete MSDR 

central statiocpower plant would involve a number ofmsystems for 

which development:work is still--required.~. It should be noted 

that some components ;such asfiumps’and control~rod drives, would 

require their own individualsystems for functions such as 

cooling -and lubric’ation.~ 1 

Given the required -ctiponents~and materials of conktruction, the 

-basic reactor $kima& and secondaj, flow systems can be designed. 

< 

However, However, the primary primary flow system would require require supporting supporting 'systems ‘systems 

'-- ‘~-dor-continuous fuel~$rocessiag,on-line fuel~processing,,"an-line fuel~analysis- and control 

--:i-mof~salt --:i-mof~salt chemistry, chemistry, reactor reactor controland-.safety, controland-.safety, handling handling of radioactive:gasesi active:gasesi fuel-=dralning--froWevery fuel-=dralning--froWevery possible possible holdup area in 

1 ->- componentCand -equipment, -afterheat control, and- temperature - 

control during-non-nuclear operations. =~ ---.. 
-42- 

The continuous fuel processing systems proposed to date are 

quite complicated and include a number of subsystems, all of 

which would have to operate satisfactorily, within the constraints 

of economics, safety, and reliability. The effects of off-design 

conditions on these systems would have to be understood so that 

control would be possible to prevent inadvertent contamination 

of the primary system by undesirable materials. 

The fuel drain system is important to both operation and safety 

since it would be used to contain the molten fuel whenever a 

need arises to drain the primary system or any component or 

instrument for maintalnence or inspection. Thus, additional 

systems would be required, each with its OWTI system for 

maintaiaing and controlling temperatures. 

tank would have-to be equipped with an auxiliary cooling Bystem 

capable of rejecting about 18 MJ(th) of heat should the need 

ariee to drain the salt immediately following nuclear operation. 

The fuel salt drain 

The secondary coolant system would also require subsystems for 

draining and controlling of salt chemistry and temperature. 

addition, the secondary loop might require systems to control 

In 

tritium and to handle the consequences of eteam generator or 

heat exchanger leaks. 
-43- 

The steam system for-the MSBR might require a departure from _ 

conventional designs hu& to the unique &Able& .associated.with 

using a coolant having a high melting temperature. Precautions 

-would have to be taken against freezing the secondary salt as it 

travels through the steam generator; suitable methods for system 

startup and control would need to be'lncorporated. ORNLhas 

proposed the &e of a supercritical steam system which operates 

at 3500 psia and provides 700°F feedwater by mixing of supercritical 

steam and high pressure feedwater. This -system would introduce 

major new development requlrem&nts because it differs from 

. 

ionveiational stesrn cycles: 

F. Makename - A Difficult Problem for the &BR 

Unlike solid fueled reactors in which the primary system contains 

act&v&ion products and only those fission products which may leak from 

< 

defective fuel pins, the *BR would have the bulk of the fission products 

dispersed throughout the reactor system. Because of this dispersal of 

. . .I __ ., 

radioactivity, remote tech&ues would be required for many maintenance 

functions if the reactor were tu h&e an acceptable plant availab'ility in 

the utility environment, 

: 

. 

'I;he MSRE was d&&&d for remote maintenance of highly radio&tive '.. 

. 

&mponents; however, no ajar roainte&&e problems -(rem&ai~ or .repair of 

_ _- 

large components) were encountered after nuclear opera&on was- initiated. 
-44- 

Thus, the degree to which the MSRE experience on maintenance is 

applicable to large commercial breeder reactors is open to question. 

As has been evident in plant layout work on nuclear~cilities to date, 

this requirement for remote maintenance will significantly affect the 

ultimate design and performance of the plant system. 

require remote techniques and tools for inspection, welding and cutting 

of pipes, mechanical assembly and disassembly of components and systems, 

and removing, transporting and handling large component items after they 

become highly radioactive. 

internals, such as graphite, might pose difficult maintenance problems 

because of the high radiation levels involved and the contamination 

The MSBR would 

The removal and replacement of core 

protection which would be requfred whenever the primary system is opened. 

Another potential problem is the afterheat generation by fission products 

which deposit in components such as the primary heat exchangers. 

Auxiliary cooling might be required to prevent damage when the fuel salt is 

drained from the primary system, and a requirement for such cooling would z 

further complicate inspection and malnteaance operations. 

In some cases, the Inspection and maintenance problems of the MSBR could 

be solved using present technology and particularly experience gained from 

fuel reprocessing plants. 

would be required in other areas, such 88 remote cuttinp., alignment, 

However, additional technology development 
7 -45- ; 

Ld cleaning and-welding of metal*members. Depending to same~-degree on the 

~particular~plant arrangement~~~ather~special tools and-equipment would 

alsohave to be-'designed 'and developed to accomplish inspection and ' 

-maintenance operations. . - ~_ 

, 

In the final-analysis, the-developinent,of adequate inspection and 

maintenance techniques and proceduresland hardware for the~MSBR< 

hinges on the success of other facets of the program, such as materials 

and~coinponent development,'and eon the~requirement that adequate care be 

taken during plant design to assure that all systems ahd components 

which would -require mainte&nce=over the life of the plant are indeed 

maintainable within the--constraints of futility operation. 

G. Safe* - Different Issues for-the MSBR :- -- -d-P 

The MS&concept-has certain characteristics which might provide 

advantages:relating to-safetym~particularly with~respect to postulated 

major types of accidents currently -considered in licensing ‘aktivities. I 

Since the fuel would be in a molten form,:cons.ideration of the core 

meltdown accident~is'not applicable to'the MSBR. Also, in the event 

of -a fuel spill;- secondary..criticality'is not-a~problem siuce-this' isathermal re'actor Zystem requiring moderator for nuclear criticality. 

_. -~. :_ .~ 

Oihei safety features'include~ the of act that Ukprimary system would 

operate at lti'pres&e'vith~fuel'salt that is more~than~lOOO'F~bel& - 
-46- 

its boiling point, that fission product iodine and strontium form 

stable compounds in the fluoride salts, and that the salts do not react 

rapidly with air or water. 

the need for excess reactivity would be decreased and some of the fission 

w 

Because of the continuous fuel processinfz, 

products would be continuously removed from the primary system. A prompt 

negative temperature coefficient of reactivity is also a characteristic 

of the fuel salt. 

Safety disadvantages, on the other hand, include the very high radioactive contamination which would be present throughout the primary 

system, fuel processing plant, and all auxiliary primary systems such 

as the fuel drain and off-gas systems. Thus, containment of these 

systems would have to be assured. 

storage systems would have to be provided by always ready and reliable 

cooling systems, particularly for the fuel drain tank and the Pa-233 decay 

Also, removal of decay heat from fuel 

tank in the reprocessing plant where megawatt quantities of decay heat 

must be removed. 

be controlled to assure safety. 

The tritium problem, already discussed, would have to I 

I 

Based on the present state of MSBR technology, it is not possible to 

provide a complete assessment of MSBR safety relative to other reactors. 

It can be stated, hawever, that the safety issues for the MSBR are 

generally different from those for solid fuel reactors, and that more 

detailed design work must be done before the safety advantages and 

disadvantages of the MSBR could be fully evaluated. 

, 
-47- 

LJ H. Codes, Standards,,~and High Temperature Design Methods 

, _ Codes and standards for MSBR equipment and systems must be developed in ~, 

conjunction with other research and development before large MSBR's can 

_ ~. 

be built. In particular, the materials of construction which are I 

currently being developed and tested would have to be certified for use 

in. nuclear power.plant applications. 

. .' 

The need for high temperature design-technology is a problem for the MSBR 

_-. .~ .~ 

as well as for other high temperature systems. The ARC currently has 

under.way a program in support of the LMFBR which is providing materials 

data and structural analysis methods for design of systems employing 

: 

various steel alloys-at temperatures up to 12OO'F. This program would ~~ .. 

need to be broade$ed to include MSBR structural materials such as 

: 

tiastelloy-N and to include temperatures as high as 1400eR to provide 

._ 

the design technology applicable3 to high-temperature, long-term _ . . .-. : :- 

operating conditions which would be expected for MSBR vessels, components, 

and core structures. 

es 
! 

-48- 

wx i VII. INDUSTRIAL PARTICXPATIOX IN THE MSBR PROGRAM 

Privately funded conceptual design studies and evaluations of HSBR 

technology were performed in 1970 by the Molten Salt Breeder Reactor 

Associates (HSBRA), a study group headed by the engineering firm of 

Black & Veatch and including five midwest utilities. 

cluded that the economic potential of the MSBR is attractive relative 

The MSBRA conto light water reactors, but they recognized a number of problems which 

must be resolved in order to realize this potential. Since that time 

the HSBRA has been relatively inactive. 

A second privately funded organization, the Molten Salt Group, is headed 

by Ebasco Services, Incorporated and includes five other industrial firms 

and fifteen utilities. 

MSBR concept and technology and concluded that existing technology is 

sufficient to justify construction of an MSBR demonstration plant 

althdugh the performance characteristics could not be predicted with 

In 1971 the Group completed an evaluation of the 

confidence. Additional support for further studies has recently been 

committed by the members of this group. 

In addition to these studies, manufacturers of graphite and Hastelloy-N 

have been cooperating vith ORNL to develop improved materials. 

I 

There has been little other industrial participation in the MSBR 

Program aside from ORNL subcontractors . At the present time, there are 
-49- 

td' two ORNL subcontracts in effect. Ebasco Services, Inc., utilizing the 

industrial, firma who are: p_articip+nta ir the Molten Salt croup is 

,_ _. 

performing a deiPfgn and evaluation -study. Foster Wheeler Corporation i8 1-~ : 

currently performing design studies on~steam generators for MSBR 

application. 

A number of factors can be identified which tend to.lbnit further 

indukrial Vinvolvement at this time, namely: 

1. The existing major industrial and utility commitment8 to the 

LWR, HTGR, and LMFBR. ., -, 

2. The lack of incentive-for industrial investment in supply& 

fuel cycle services, , such as those required for solid fire1 

rektors. 

, 3. The overwhelming manufacturing and operating experience with 

solid fuel reactor8 in contrast with the very limited hvolve- _ 

merit with fluid fueled reactors. 

4. The less advanced state of MSBR technology and the lack of 

deknstrated solutious to the major technical problems ~\ 

associated with the MSBR concept. 
-50- 

It should be noted that these factors are also relevant considerations 

in establishing the level of governmental support for the MSBR program 

which in turn, to some extent, affects 

and utility industries. 

the interest of the manufacturing 
-51- 

VK CONCLtiSIONS 

The Molten Salt Breeder Reactor, if successfully developed and marketed, 

'&uld provide a useful supplement to the currently developing uraniumplutonium reactor economy. This concept offers the potential for: 

. Breeding in a thermal spectrum reactor; 

. Efficient use of thorium as a fertile~material; 

", i 

. Elimination.of 'fuel fabrication and spent fuel shippink; 

. High thermal efficiencies. 

Notwithstanding these attractive features, this assessment has ,' 

reconfirmed the existence of major technological and engineering 

problems affecting feasibility of the'concept as a reliable and ~-' : 'i -~- 

economic breeder for the utility industry. The principal concerns _. 

include uncertainties with materials, with methods of controlling I 

..~ _ 

;- 

tritium, and with the design of components and systems along with 

, - , 

their special handling, inspection and maintenance equipment. Many 

of thes'e problems are &mpounded by the use of a fluid fuel in which ~- 

fission products and delayed neutrons are distributed throughout the 

primary reactor and reprocessing systems. 

_ 

The resolution of the problems of the MSBR will require the conduct \ 

of'an IntkUive research and development program. -kcluded among 

_ 

the major effork that &&la have to bt? accompiishedare: 

. _. ~I 

. Proof testing.of an integrkged reprocessing system; 

: ; - * 

. Development of asuitable containment material; 
-52- 

. Development of a satisfactory method for the control and 

retention of tritium; 

L4 _I 

. 

. 

. 

Attainment of a thorough understanding of the behavior of fission 

products in a molten salt system; 

Development of long' life moderator graphite, suitable for 

breeder ' applica - 

. Conceptual definition of the engineering features of the many 

components and sys tems ; 

. Development of adequate methods and equipment for remote 

inspection, handling, and maintenance of the plant. 

- 

- 

The major problems associated with the MSBR are rather difficult In nature 

and many are unique to this concept. Continuing support of the research 

and development effort will be required t6 obtain satisfactory solutions 

to the problems. When si~q~ificant evidence is available that demonstrates 

realistic solutions are practical, a further assessment could then be made 

as to the advisability of crdvancini into 

engineering phase of the development process inclu+g that of industrial 

involvement. Proceeding with this next step would also be contingent 

upon obtaining a firm demonstration of interest and commitment to the 

concept by the power industry and the utilities and reasonable assurances 

that large scale government and industrial resources can be made avail 

- 

on a continuing basis to this program in light of other canrmitments to 

the commercial nuclear power program and higher priority energy 

development efforts . 
-53- 

u 

. 

p. REFERENCES _ 

1. U. S. Atomic l%uergy'Conunission, "The 1967~Supplement to the 1962‘ 

Rep&t to the Preaidkt on Civilian Nuclear Power" USMC Repoit, 

- February 1967. 

2. U. S. Atomic Energy -&nrkk&m, %he ~&x-of Tho&un%a Nuclear 

Power Reactom" USA&-Rep& WASH-1097, 1963.' -A ' 

” 

, 3. U. S. Atomic Ene&y Comaiissibn, "P;'otentkl Nuclear Power Gro&h 

-Patterns," US&i% -Rep& tiASiGlCJ98, Decder 1970. 

'4. U. S. Atomic Energy Cormission~ "Cost-Benefit &naly& of- the 

U. S. Breeder ~R&&or%ogrk;" USAEC'keiort GASR-1126,“1969. ' 

5. U. S. Atomic Energy Comi&Lss&n~ "Updateid- (1970) Co&B&efit Analysis 

of the U. S. Breeder Reactor Program,' USAEC Report WASH-1184, 

. 

January 1972.. ~~ 

,. 

.- 

.- 

6. Edison Electric'ktitute, "'&port bn the EEf Reactor Assessment Panel," 

EEI Publication No. 70-30, 1970. 

7. AnntkHekngs on &i&tor$ev~lo~m&t P~o&.m, U. S.'At&ic Energy 

" &mmisS~ion h 1972~Aut&i&g '%&$lation; Hear&&s before -&Joint 

7~ 

Committee on Atomic Energy, Congress of the United-States p. 820-830, 

P. 

Molten salt Breeder Reactor,' ORAL-4541, June 1971.' -_ . .

\ 

. 
-54- 

10 . 

11. 

12 . 

13. 

14 . 

16 . 

17. 

Roaenthal, M. I?., et al.; "Advances in the Development of Molten-Salt 

Breeder Reactor8," A/CONF. 49/P-048, Fourth United Nations 

International Conference 011 the Peaceful Uses of Atomic Energy, 

Geneva, September 6-16, 1971. 

Trinko, J. R. (ea.), "Molten-Salt Reactor Technology," Technical 

Report of the Molten-Salt Group, Part I, December 1971. 

Trinko, J. R. (ed), "Evaluation of a 1000 We Molten-Salt Breeder 

Reactor," Technical Report of the Molten Salt Croup, Part 11, 

Navemb er 19 7 1 . Ebasco Seroices Iuc., "1000 MWe Molten Salt Breeder Reactor Conceotual 

W 

Design Study," Final Report Task I, Prepared under ORNL subcontract 

3560, February 1972. 

"Project for Investigation of Molten Salt Breeder Reactor,'' Final 

Report, Phase I Study for Molten Salt Breeder Reactor Associates, 

Sepfember 1970. 

Cardwell, D. W. and Haubenreich, P. N., "Indexed Abstracts of 

Selected References on Molten-Salt Reactor Technology," ORNL-TM-3595, 

December 1971. 

Kasten, P. R., Bettfs, E. S. and Robertson, R. C. , "Design Studies 

of 1000 MW(e) Molten-Salt Bkeder Reactors ," ORNL-3996, August 1966 . Molten Salt Reactor Progr'am SPm$annual Reports beginning in 

February 1962. 
i 

A- 1 

i 

f 

- General 

Thermal capacity of reactor 2250 MW(t) 2250 MW(tj 

Gross electrical generation 1035 MW(e) 1035 MW(e) 

Net 1000 MW(e) 1000 MW(e) electrical output Net overall thermal efficiency 44.4% 44.4% i 

Net plant heat rate 7690 Btu/kWhr 2252 J/kW-SC . i 

1 

Structures 

Reactor cell, diameter X height 

Confinement building, diameter X height 

Vessel ID 6.77 m 

Vessel height at center (approx) 20 ft 6.1 m 

Vessel wall thickness 2 in. 5.08 cm 

Vessel head thickness 3 in. 7.62 cm 

Vessel design pressure (abs) 75 psi 5.2 X los N/mz 

Core height . 13ft ' 3.96 m 

Radial thickness of reflector 30 in. 0.762 m 

Volume fraction of salt in central core zone 0.1 3 

Volume fraction of salt in outer core zone 0.37 0.37 

Average overall core power density 22.2 kW/liter 22.2 kW/liter . 

Peak power density in core 70.4 kW/liter 70.4 kW/liter 

Average thermal-neutron flux 2.6 X 1014 neuGons cm-' sec-' 2.6 X 1014 neutrons scc-' 

Peak thermal-neutron flux 8.3 X lo'* neutrons sec-' 8.3 x neutrons cm-l scc-l 

Maximum graphite damage flux (>SO keV) 3.5 x 1014 neutronscrnm2 scc-l 

Damage flux at. maximum damage 3.3 x 1014 neutroks cm-3 sw-' 

Graphite temperature at max 1284' F 969°K 

Graphite temperature at maximum graphite 1307" F 982°K 

72 x 42 ft 

134 X 189 ft 

22.0 X 12.8 m 

40.8 X 57.6 m 

Reactor 

Number of core elements > 1412 1412 i 

0.13 

3.5 x 10" neutrons cm" sec-' 

3.3 x 1014 neutrons cm-? set" i region (approx) 

flux region , 

damage region 

Estimated useful life of graphite 4 years 4 years ,i 

i 

Maximum flow velocity of salt in core 8.5 fps 2.6 m/xc i 

Total fuel salt in reactor vessel 1074 ft3 30.4 m3 i 

Fissile-fuel inventory in reactor primary 3316 Ib i 

Total weight of graphite in reactor 669,000 Ib 304,OOOkg .e 

Total fuel-salt volume in primary system 1720 ft3 48.7 m3 

, system and fuel processing plant 

Thorium inventory I50.000 Ib 

Breeding ratio 1.06 1.06 

1 Yield 3.2 %/year 3.2 %/year 

Doubling time, compounded continuously. 22 years 22 years 

at 80% power factor 

Primary heat exchangers (for each 

Thermal capacity. each 

Tube-side conditions (fuel salt) 

\ 

Tube OD 

Tube length (approx) 

Number of tubes 

Inlet-outlet conditions 

Mass flow rate 

Total heat transfer surface c 

Shell ID 1.73 m 

Mass flow rate 17.6 X lo6 lb/hr 2218 kg/sec 

Shell-side conditions (coolant salt) ,- 

Inlet-outlet temperatures 727-894" K 

Overall heat transfer coefficient (approx) 850 Btu lu" ft-2 (OF)-' 4820 W m-2 (OK)-@ 

U 

3 
A-2 

Appendix A (continued) 

Engineering unit? International systcm.units* 

Primary pumps (for each of 4 units) 

Pump capacity, nominal 16,000 gpm 1-01 m'/sec 

Rated head 150 ft 45.7 m 

Speed 890 rpm . 93.2 radianshc 

Specific speed 2625 rpm(gp~n)~'~/(ft)~*~~ 5.321 radian~/sec(m~~sec)~.~/(m)~~~~ 

impeller input power 2350 hp 1752 kW 

Design temperature 1300" F 978°K 

. Pump capacity, nominal 20,000 gpm 1.262 m3/sec 

Rated head 300 ft 91.4 m 

Speed, principal 1190rpm 124.6 radianJsec 

Specific speed 

Ihpcller input power 3100 hp 2310 kW 

Design temperature 1300°F 978°K 

Outside diameter 14 ft 4.27 m 

Overall height 22 ft 6.71 m Storage capacity 2500 ft3 70.8 m3 

Design pressure 55 psi 3.79 X los N/m2 

Number of coolant U-tubes 1500 1 500 

Sue of tubes, OD 9' in. 1.91 cm 

Number of separate coolant circuits 

Coolant fluid 'LiF-BeFs 

Under normal steady-state conditions: 

Secondary pumps (for each of 4 units) 

2330 rpm(gpm)0'5/(ft)0'7s 4.73 radianJsec(m' /sec)O 4 /( m)O *7s 

Fuel-salt drain tank (1 unit) 

40 

Maximum heat load 18 MW(t) 18 MW(t) 

Coolant circulation rate 830 gpm 0.0524 m3/sec 

Maximum tank wall temperature -1260°F -955°K 

Coolant temperatures, in/out 900-1050°F 755-839°K 

Maximum transient heat load 53 Mw(t) 53 MW(t) 

Fuel-salt storage tank (1 unit) 

Storage capacity 2sOOft3 - 70.8 m' 

Coolant fluid Boiling water . Boilingwater 

Total volume of coolant salt in systems 

Storage capacity of each tank 2100 ft3 59.5 m' 

Heat-removal capacity. fmt tank in series 400 Irw 400 kW 

Thermal capacity 120.7 MU'( t) 120.7 MW(t) 

Tube-side conditions (steam at 3600-3800 

- Heat-removal capacity 1 MWW 1 MW(t) 

Coolant-salt storage tanks (4 units) 

8400 ft' 237.9 m3 

Steam generators (for each of 16 units) 

psi) 

Tube OD y in. 1.27 cm 

Tube-sheet-to-tube-sheet length (approx) 76.4 ft 23.3 m 

Number of tubes 393 393 

I nle tbutle t temperatures 700- 1000" F 644-811°K - Mass flow rate 633,000 lbbr 79.76 kg/sec 

Total heat transfer surface 3929 ft2 365m' . 

Shell ID 1.5 ft . 0.457 m 

Inlet-outlet temperatures 1150-850°F 894-727% 

Mass flow rate 3.82 X lo6 lb/hr 481.3 Wsec 

range 490-530 Btu hr" 11- CF)" 2780-3005 W m-2 ("K)'t 

Shell-side conditions (coolant salt) . 

Apparent overall heat transfer coeffKient 
A-3 

Appendix A (cont hued) 

.~ 

Engineering unit.@ International system unitsb 

Steam reheaters (for each of 8 units) 

Thermal capacity . 36.6 W(t) 36.6 MW(t) 

Tube-side conditions (steam at 5sO psi) 

Tube OD 36 in. 1.9 cm 

Tube length 30.3 ft 9.24 m Number of tubes 400 400 

Mass flow rate 

Total heat transfer surface 2381 Et2 221.2 m2 

Shell ID 21.2 in. 0.54 m 

Mass flow rate 1.16 X IO6 Ib/hr 146.2 kg/sec 

Inlet-outlet temperatures 650- 1000” F 616-811°K 

64 1,000 Ib/hr 80.77 kg/sec 

Shell-side conditions (coolant salt) 

Inlet-outlet temperatures 1150-850°F 894-727°K 

Overall heat transfer coefficient 298 Btu hr’’ ft’) C 1690 W m-2 (OK)-* 

Turbine-generator plant (see “General” above) 

Number of turbme-generator units 1 

3500 psia. 1000°F 

7.15 X lo6 lb/hr 

540 psia. 1000°F 

Turbine 24.1 X IO6 N/m2, 811°K throttle conditions 

Turbine 900.9 kg/sec throttle mass flow rate 

Reheat steam to IP turbine 3.72 X lo6 N/m2. 81 1°K 

Condensing pressure (abs) 1.5 in. Hg 5,078 N/m2 

Boiler feed pump work 19.700 hp 14,690 kW 

Booster feed pump work (motordriven), - 6200hp 4620 kW 

(steamturbine-driven), each of 2 units 

each of 2 units 

Fuehalt inventory, primary system 

Reactor 

Core zone 1 290 ft’ 8.2 d 

Core zone 11 382 ft’ 10.8 m5 

Plenums, inlets, outlets 218 ft’ 6.2 m3 

2-in. annulus 135 ft’ 3.8 m’ 

Reflectors 49 ft’ 1.4 m3 

Tubes 269 ft’ 7.6 m’ 

Inlets, outlets 27 ft3 0.8 m’ 

Pump bowls 185 ft3 5.2 m’ 

145 ft’ 4.1 m3 

0.3 m3 

Piping, including drain line 

Off-gas bypass loop 

Tank heels and miscellane 0.3 m3 

48.7 m3 

Primary heat exchangers 

Total enriched salt in primary system 

- 

Fuet-processing system (Chemical Treatment 

Cyde time for salt inventory 

Heat generation in sa!t to processing plant 

Components 

Compositioh 

Molecular weight (approx) . Melting temperature (approx) 

Vapor pressure at 1150°F (894.3%) 

propertie 

Wan’) = 3.752 - 6.68 x 10*r 

P (Uft’) = 235.0 - 0,02317t(%“) 

At 13WF (978°K) 204.9 Ib/ft’ 3283.9 Wm3 

At 1175°F (908°K) 207.8 lb/ft3 

At 1050°F (837K) 210.7 lb/ft3 

LJ 
A-4 .- 

Appendix A (continued) 

Engineering unitP International system unit# 

Viscosity:d p (centipoises) = 0.109 

exp [4090/T('K)];p (Ib it-' h-l) 

= 0.2637 exp [7362/T("R)] 

At 1300°F (978°K) 

At 1 175°F (908°K) 

At 1050°F (839°K) 

17.3 Ib hr-' ft-' 

23.8 lb hr-' ft" 

34.5 Ib hr-' ft-' 

0.324 Btu Ib" PF)" * 4% 

0.69 Btu hr-' Po-' ft" 

0.71 Btu hr" ("F)" ft-' 

0.69 Btu hr" ("F)-' ft-' 

0.007 N sec m-2 

0.010 N sec rn3 

0.015 N sec m-2 

1.357 J g" PIC)'' * 4% 

1.19 W m-' CK)-' 

1.23 W m-' VK)" 

1.19 W m-' PK)" 

Heat capacitf (specific heat, cp) 

Thermal conductivity (k)f 

At 1300°F (978°K) 

At 1175°F (908°K) 

At 1050°F (839°K) 

Design properties of cootant salt 

Com ponen t s NaBF4-NaF NaBF4-NaF 

Composition 92-8 mole % 92-8 mole % 

Melting temperature (approx) 725°F 658°K 

Vapor pressure:g log P (mm Hg) 

Molecular weight (approx) 104 104 

= 9.024 - 5920/T("K) 

At 850°F (727°K) 8 mm Hg 1066 N/m2 

At 1150°F (894°K) 252 mm Hg 33,580 N/m2 

("C); p Curt3) = 141.4 - 0.0247r (OF) 

At 1150°F (894°K) 11 3.0 lb/ft3 181 1.1 kg/m3 

At 1000"F(81l0K) 116.7 Wft3 1870.4 kg/m3 

At SSO" F (727" K) 120.4 lblft' 1929.7 kg/m3 

exp [2240/T (OK)] ; p (Ib, ft" hr-l) 

= 0.2121 exp [4032/T('R)) 

At 1150"F(894°K) 

At 1000°F (811°K) 

At 850°F (727°K) 

Density:=p (g/cm3) = 2.252 - 7.1 1 X 104t 

Viscosity:d p (centipoises) = 0.0877 

, 

2.6 Ib ft-' hr" 

3.4 lb ft-' hr-' 

4.6 lb ft-' hr-' 

0.360 Btu lb-' (OF)" * 2% 

0.23 Btu hr" (OF)" ft-' 

0.23 Btu hr" (OF)" ft" 

0.26 Btu hr" (OF)" ft" 

0.001 1 N sec m-2 

0.0014 N sec in-2 

0.0019 N sec m-2 

1507 J kg-' (OK)-' * 2% 

0.398 W m-l (OK)" 

0.398 W an-' (OK)-' 

0.450 W m-' (OK)-' 

Heat capacityh (specific heat, cp) 

Thermal conductivity (k)' 

At 1150"F(894°K) 

At 1M)O"F (811OK) 

At 850°F (727°K) 

Design properties of graphitei 

Density. at 70°F (294.YK) 115 lb/ft3 1843 kg/m3 

Bending strength 4000-6000 psi 28 X 106-41 X lo6 N/m2 

Modulus of elasticity coefficient 1.7 X lo6 psi 

Poisson's ratio 0.27 0.27 

Thermal expansion coefficient 2.3 X 10df'F 1.3 X 10dPK 

Thermal conductivity at 1200°F. ' 18 Btu hr-' (OF)" ft-' 31.2 W m-I (OK)" 

Electrical resistivity 

Specific heat 

11.7 X lo9 N/m2 

unirradiated (approx) 

8.9 X 104-9.9X lo4 n-cm 

0:33 Btu lb-' (OF)-' 

1 x 10" cm2/sec 

8.9 X 104-9.9 X lo4 ntm 

At 600°F (588.8"K) 1380 J kg" (OK)" 

At l2WF (922.0"K) 0.42 Btu lb" (OF)-' . 1760 J kp-' (OK)" 

1 x IO* cm2/sec 

surfaces 

Helium permeability at STP with sealed 

Bd 
A-5 

Appendix A (continued) 

Engineering unitP International system unitsb 

Design properties of Hastelloy Nk 

Density 

At 80°F (300°K) 

At 1300°F (978°K) 

Thermal conductivity 

At 80°F (300°K) 

At 1300°F (978°K) 

At 80°F (300°K) 

At 1300°F (978°K) 

Thermal expansion 

At 80°F (300°K) 

At 1300"F(978"K) 

At 80°F (300°K) 

At 1300°F (978°K) 

Tensile strength (approx) 

At 80°F (300°K) 

At 1300°F (978°K) 

Specific heat 

Modulus of elasticity coefficient 

557 lblft3 

541 lb/ft' 

6.0 Btu hr-' CF)" ft-' 

I 12.6 Btuhr" ("F)" ft-' 

. 0.098 Btu lb-' (OF)-' 

0.136 Btu lb-' (OF)" 

5.7 x IO'PF 

9.5 x 10"PF 

31 X lo6 psi 

25 X lo6 psi 

115,000 psi 

75.000 psi 

8927 kg/m3 

8671 kg/m3 

10.4 W m-' (OK)-' 

21.8 W m-I (OK)-' 

410 J kp-' ("IC)-' 

569 J kg-' (OK)-' 

3.2X 1O"PK 

5.3 X 10df'K 

214 X lo9 N/m2 

172 X lo9 N/m2 

793X 106N/m2 

517 X lo6 N/m2 

Maximum allowable design stress - 

C t 80" F (300°K) 

At 1300°F (978°K) 3500 psi 24 X lo6 N/m2 

25,000 psi 172 X IO6 N/m2 

Melting temperature 2500" F ' 1644°K 

'English engineering units as used in MSR literature. 

bMeter-kilogram-second system. Table closely follows International System (SI). See Appendix C for conversion factors from 

- engineering to SI units.