ORNL-TM-3177.txt

 

ORNL-TM-3177

Contract No. W-ThO5-eng-26

Reactor Division

MOLTEN SALT BREEDER EXPERIMENT DESIGN BASES

J. R. McWherter

—— - LEGAL NOTICE
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 of implied, or assumes any
legal liability or responsibility for the accuracy, com-
pleteness or usefulness of any information, apparatus,
product or process disclosed, or represents that its use
would not infringe privately owned rights. J

 

 

 

NOVEMBER 1970

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION

RISTIIBUTION OF THIS DOCUMENT IS UNLIMITRL

e

5
iii

CONTENTS
List of Tables . ¢ o o o o o o o o o o o o
List of Figures.
Acknowledgment .
Abstract .
1. Introduction . . ¢« « « ¢« ¢« & « o « &
2. The MSRE and the Present Status of the Technology.
3. Reference Plant Design - MSER.
4. MSBE Design Bases.

4.1 MSBE Requirements
4.2 WMSBE Core and Pressure Vessel Configuration .
4.3 MSBE Primary Salt Properties.
L.L MSBE Primary System .
L.4.1 Primary Loop .
4.4.1.1 Primary Pump.
L. 4.,1.2 Primary Heat Exchanger.
4.4.2 Gas Separation Bypass.
4h.4.3 Primary Salt Drain Tank.
4. 4.4t Primary Salt Storage Tank.
4.4.5 Primary Salt Sample System .
4.5 MSBE Secondary System . « + « o« + « +
4.6 MSBE Steam System .
4.7 MSBE Reactor Cell .

4.8 Drain Tank Cell, Off-Gas Cell, and Secondary Cell .

4.9 MSBE Reactor Building . . . . . . .
4,10 MSBE Chemical Processing. « « « « o« « o o 4 &
MSBE Expected Accomplishments.
Appendix I - MSRE.
6.1 Description .
6.2 Experience.
6.2.1 Fuel Chemistry .
6.2.2 Materials.

vii

CoO VYUl VO Y

16
16
16
18
18
20
20
21
21
2P
23
23
26
26
27
29
30
30
35
36
38
iv

6.2.3 Nuclear .
6.2.4 Equipment .

6.2.5 Maintenance of Radioactive Systems.

7. Appendix IT .
7.1 MSBR Plant Description . . . . .« « « « &

ReferencCes « o o o o s o o o o o o s o s o o »
Table I.
Table II.

Table IIT.

LIST OF TABLES

Comparison of Design Data, MSRE, MSBE, MSER .

Nuclear Characteristics of Several Conceptual
MSBE Reactor Configurations .

MSBE Irradiation Facility V-1 Neutron Flux.

1h
Figure
Figure
Figure
Figure
Figure
Figure
Figure
Figure
Figure
Figure
Figure
Figure
Figure
Figure

Figure

O O 1 Oy V1 WO

o R R
Vi W D = O

vi

LIST OF FIGURES

MSBE Reactor Core Assembly .

MSBE Core Section Plan .

MSBE Flow Diagram.

MSBE Primary Heat Exchanger.

MSBE Reactor Building Section B-B.
MSBE Reactor Building Plan A-A . . .
MSBE Chemical Processing Flowsheet .
MSRE Flowsheet .

Layout of the MSRE .

. Flow Diagram for MSBR Plant.

. Plan View of MSBR at Reactor Cell Elevation.
. Sectional Elevation Through MSBR Plant Building.
. BElevation of MSBR Drain Tank Cell.

. Plan View of MSBR Vessel .

. Sectional Elevation of MSBR Vessel .
vii’
ACKNOWLEDGMENT

The author gratefully acknowledges the layout design work by W. Terry
for this report and the general contributions to the report by R. B. Briggs,
P. N. Haubenreich, M. I. Lundin, H. E. McCoy and L. E. McNeese. The
reactor physics calculations in support of this report were made by O. L.
Smith, W. R. Cobb, and J. H. Carswell. The general concept is based on
studies of the 1000 Mw(e) single-fluid Molten Salt Breeder Reactor by
E. S. Bettis, C. W. Collins, W. K. Furlong, E. C. Hise, H. A. Mclain,

H. M. Poly, D. Scott, H. L. Watts, and others.
MOLTEN SALT BREEDER EXPERIMENT DESIGN BASES

J. R. McWherter

ABSTRACT

The design bases for the Molten Salt Breeder Experi-
ment (MSBE) are based on information from the MSRE and
the reference plant design of a 1000 Mw(e) single-fluid
Molten Salt Breeder Reactor (MSER).

Calculations indicate that a 150 Mw(thermal) reactor
is a reasonable size that would meet the project objectives
for the MSEBE.

The primary salt for the MSBE contains both the
fissile (®33U) and the fertile (Th) material. The heat
generated in the primary system is transferred by a
secondary salt loop to the steam generators.

Provisions are made in the MSBE core to permit ex-
posure of removable graphite samples at conditions similar
to those expected in the MSBR.

The pumps and heat exchangers in the MSBE are similar
to those proposed for the MSBR.

 

Keywords: design, design criteria, design data, experiment,
fluid-fueled reactors, fused salts, graphite, MSBE, MSER,
reactors.

1. INTRODUCTION

The goal of the Molten Salt Reactor Program (MSRP) is to provide
the basic scientific and engineering data and the experience necessary
for the development and construction of large molten salt reactors to
produce economical electricity. ©Such reactors would be based on the use
of fluoride salts containing dissolved fissionable material (238U, 238y
or plutonium) and fertile material (thorium). We believe that in the
long run the most economical embodiment of the molten salt reactor con-
cept will be a reactor with a positive breeding gain. ©Since an essential
requirement for such a breeder will be a salt processing facility closely
coupled to the reactor system, the MSRP embraces processing as well as

reactor development.
A major step in the program was the construction and successful
operation of the Molten Salt Reactor Experiment (MSRE). The MSRE was
a circulating molten-salt-fueled, graphite-moderated reactor that operated
at 7.3 Mw(thermal) and a core outlet temperature of 1210°F. The purpose
of the MSRE was to provide a demonstration of the technology as it existed
in the early 1960's and a facility for investigating the compatibility
of fuels and materials, the chemistry of the fuels, and the engineering
Teatures of molten-salt reactors.

Four years of MSRE operation provided an essential base for proceed-
ing with larger reactors. However, the MSRE was a small reactor with a
low power density and contained no thorium in the salt. We believe that
one step that is highly desirable before building a prototype power
breeder plant is the construction of a reactor with a power density near
that of the larger reactors, with a fuel composition like that of a power
breeder, and which will produce protactinium at a sufficient rate to pro-
vide for processing development. This Molten Salt Breeder Experiment
(MSBE) should include the essential features of a power breeder and satisfy
as many of the technical criteria of the reference design as practical.
The size and power of the MSBE should be no greater than will be necessary
to meet these requirements. The experiment would demonstrate all the
basic equipment and processes at proposed design conditions of the large
plants. The essential purpose of the MSBE would be to produce information
rather than electricity, but it should demonstrate the technology of the
production of steam at conditions adequate for electrical production.

This report expands on these design bases for the MSBE.

2. THE MSRE AND THE PRESENT STATUS OF THE TECHNOLOGY

Molten salt reactor technology has been under development since
1947, with the most prominent accomplishments being the operation of the
Aircraft Reactor Experiment! in 1954 and the MSRE from 1964 to 1969.
Much of the present status of the technology is best described in terms
of the MSRE. OSome of the important characteristics of that reactor are
given in Table I. The MSRE information in Table I is in general from

Refs. 2 and 3; however, it has been modified as required to reflect the
Table I. Comparison of Design Data, MSRE, MSBE, MSBR

 

 

MSRE MSEE MSBRE
Reactor power, Mw(t) 7.3 150 2250
Breeding ratio - 0.96 1.06
Peak graphite damage flux, x 1012 5 x 1014 3 x 104
(E. > 50 kev)
n
neutrons/cm? *sec
Peak power density, w/cc
Primary salt 30 760 500
Core including graphite 6.6 114 65
Peak neutron heating in 0.2 2.6 1.7
graphite, w/cm®
Peak gamma heating in 0.7 6.3 .7
graphite, w/cm®
Volume-fraction primary salt in core 0.225 0.15 0.13
Composition, mole %
LiF 65 71.5 T1.7
BeF, 29.1 16 16
ThF, None 12 12
UF, 0.9 0.5 0.3
ZxFy 5 None None
Liquidus, °F 813 932 932
Density, 1b,/ft® at 1100°F 141 2112 210
Viscosity, 1b/ft-hr at 1100°F 19 ogb 29
Heat capacity, Btu/1lb,°F 0.47 0.32 0.32
Thermal conductivity, 0.83 0.75 0.75
Btu/hr-ft-°F
Volumetric heat capacity, 66 66 66
Btu/ft2:°F
Temperature, °F
Inlet reactor vessel 1170 1050 1050
Outlet reactor vessel 1210 1300 1300
Circulating primary salt vol, ft2 T0 266 1720
Inventory fissile, kg 32¢ 396° 1470
Power density primary salt L 20 L6

circulating average, w/cc

 

%206 @ 1300°F; 212 @ 1050°F.
P16.4 @ 1300°F; 34.2 @ 1050°F.

C
2830 initial.
(continued)

 

 

MSRE MSEBE MSBR®
Number of primary loops 1 1 b
Primary pump capacity, gpm 1200 5400% 16,000
Secondary system salt LiF-Bel, NaBF, -NaF  NaBF, -NaF
Composition, mole % 66 - 34 92 - 8 92 - 8
Liquidus temp, °F 850 25 25
Density, 1b,/ft® @ 1000°F 12k.1 117 117
Viscosity, lby/ft-hr @ 1000°F 28.7 3.4 3.4
Heat capacity, Btu/lb,°F 0.57 0.36 0.36
Thermal conductivity 0.58 0.27 0.27
Btu/hr-ft-°F
Temperature, °F
Heat exchanger inlet 1015 850 850
Heat exchanger outlet 1075 1150 1150
Number of secondary pumps 1 1 Iy
Secondary pump capacity, gpm 850 5300 20,000
Tertiary system air steam steam
Tnlet temp., °F ~ 70 700 700
Outlet temp., °F ~ 180 1000 1000
Outlet pressure, psia b7 3600 3600

 

YFor 200°F AT; 4300 gpm required at 250°F AT.

measurements made during operation of the reactor.4

A description of

the MSRE and a summary of the experience with it are given in Appendix I.

Some significant advances in the technology have been made with

materials since the construction of the MSRE. The alloy, Hastelloy N,

which was used throughout the MSRE salt systems, was modified to improve

its resistance to damage by neutron irradiation while retaining its

excellent corrosion resistance.

Graphites have been made which change

dimensions less at a given neutron dose and temperature than that graphite

used in the MSRE, and methods have been devised to seal the graphite pores

to reduce gas permeability and therefore xenon poisoning in high-flux

reactors.

Since the inception of the MSRE, great strides have been made in

salt processing, notably the invention of the reductive extraction and
metal transport processes. The basic chemistry has been shown to be quite
favorable. Materials of construction for the process eguipment are under
development, with encouraging progress being made. These new processes
make a single fluid reactor concept similar to the MSRE, but with a higher
power density, a potential high performance breeder. By permitting a
breeder core to resemble a scaled-up MSRE, these developments have made
the MSRE technology more directly applicable.

Another advance since the MSRE has been the experience with molten
sodium flucroborate salt, a lower-melting and less expensive coolant than

was used in the MSRE secondary system.

3. REFERENCE PLANT DESIGN - MSBR

A conceptual design of a single-fluid 1000 Mw(e) Molten Salt Breeder
Reactor (MSBR) power station was made and is described in Appendix IT
and in more detaill elsewhere.®® The fuel is contained in the primary salt
which 1s a mixture of fluorides containing also the fertile material.
Some basic design conditions are given in Table I. The MSBE design is
based on this MSBR concept, including current revisions, insofar as is

practical.
L. MSBE DESIGN BASES

4.1 MSBE Requirements

 

The MSBE should demonstrate the basic technology of a large molten
salt breeder reactor so that moderate scale-up and normal improvement
of equipment and processes are the major requirements for building large
plants. The plant should be as small as 1s consistent with making a
complete demonstration.

Major criteria for the plant are the following:

1. The plant shall be a facility for testing materials, components,
systems, and methods at conditions, where practical, equal to or more
severe than those of the reference MSBR. For instance, it is desired
that the damage neutron flux in the MSBE graphite be the maximum that

has been proposed in molten salt reactor studies, which is about twice
that proposed for the reference MSBR. In other cases, such as the average
circulating power density in the salt, it may not be practical to egual
that proposed for the reference MSBR.

2. The reactor shall have the capability for exposing to a fast
(> 50 kev) neutron flux of 5 x 10'% neut. cm™@ sec™! in primary salt at
temperatures up to 1300°F core graphite elements that are of full MSBR
cross section. These elements may be shorter in length than those pro-
posed for the MSBR. At least one of these elements shall be individually
removable. In this region of the core, the power density shall be agbove
500 w/ce of salt and the salt flow conditions shall be as close as prac-
tical to those proposed for the MSBR. This will permit evaluation of the
useful life of potential MSBR graphite at irradiation and thermal condi-
tions equal to or more severe than those proposed for the MSBR.

3. The primary and secondary salt compositions shall be essentially
the same as proposed for the single fluid MSBR. Modifications shall be
limited to those which will not significantly alter the chemistry or physi-
cal properties of the salts. This will permit studies of the nuclear and
chemical effects in the salt, heat transfer and fluid flow characteristics,
and the chemical processing aspects at conditions as near those of the
proposed MSBR as practical.

L. The design power of the reactor shall be sufficient to meet the
above criteria and in addition supply considerable protactinium and fission
proaucts for process development. A conversion ratio near 1.0 is desirable
but not essential. In addition, the average power density of the circu-
lating primary salt shall be near that of the MSBR (46 w/cc). This will
permit some determination of the fission product handling problems, such
as afterheat, in the fluid processing systems.

5. The design of the plant shall be similar to that proposed for
the MSBR. Where practical, the MSBE primary system components shall be
similar in design to those proposed for the MSBR with a design life of
thirty years and of a size that can be scaled up for use in demonstration
plants. The flowsheets of the two plants shall be similar where practical.
With this approach the design and operation of the MSBE will give con-

siderable advance information for design and operation of the MSBR.
6. The maximum operating temperatures, and, where practical, the
temperature differences in the MSBE shall be the same as those of the
MSBR. This should permit evaluation of materials and systems at MSER
thermal conditions.

7. Thermal energy shall be transferred from the primary salt to a
secondary salt from which it shall be removed by steam generation at pro-
posed MSBR conditions. This provides a double barrier between the fission
products and the steam system and provides experience with steam genera-
tors, a major undeveloped component for molten salt reactors.

8. The generation of electricity will only be required if it is
economically Justified, since the effect of the steam system operation
on the nuclear systems can be determined without any specific use of the
steam.

9. The chemical processing of the primary salt shall be done by
processes proposed for the MSBR and with equipment similar.to that
for the MSBR. This requirement'stems from the importance of fuel pro-
cessing to an MSBR and the need to experiment with and demonstrate the
process on highly irradiated salt. The MSBE will be an excellent source
of irradiated primary salt for use in evaluation of proposed chemical
processing schemes and equipment.

10. Maintenance techniques and procedures proposed for the MSER,
including removal and replacement of the core graphite, shall be used
where practical in the MSBE. This will permit development of maintenance
techniques and procedures under conditions similar to those proposed for
the MSBR.

11. In support of the above requirements and to improve the under-
standing of molten salt reactor systems, facilities shall be provided for:

(a) on-line chemical analyses of the fuel salt (for the most impor-

tant constituents),

(b) obtéining unbiased samples of the salts for complete chemical

and isotopic analyses,

(¢) determining compatibility of materials by examination of

removable specimens, '

(d) studying the composition of gas at various locations,
(e) studying the deposition of fission products,
(f) determining the behavior of tritium,
(g) removing the core graphite array for post-irradiation examination,
(h) continuously monitoring the nuclear reactivity,
(i) determining the dynamic characteristics of the entire system,
(j) examining reactor components after operation with irradiated
salt,

(k) monitoring the behavior of components.

L.2 MSBE Core and Pressure Vessel Configuration

Preliminary calculations?®® were made by 0. L. Smith, W. R. Cobb,
and J. E. Carswell of small-single-fluid reactors to determine the breed-
ing ratio and the power required in various configurations of core and
blanket to achieve a peak damage flux of 5 x 10'%* neutrons cm™2 sec~!

(En > 50 kev). The core, of course, was composed of graphite and salt.

A range of salt fractions in the core between 0.1 and 0.2 was studied.
The blanket was 100% salt in the radial direction but was assumed to con-
tain 30 to 50% graphite in the axial direction. The axial plena at the
top and bottom of the reactor vessel were assumed to contain 9&% salt
and 6% Hastelloy N for structural purposes.

The salt composition range considered was (in mole percent):

16-20% BeF,, 12-14% ThF,, and the balance as 7LiF with sufficient 233UF,
for criticality. Of course, 235UF, could be used, if desired.

The results of some of these initial calculationg are given in
Table II. The remainder of the cases are reported elsewhere.”>®® The
breeding ratio reported is the value at start-of-life conditions, assuming
pure 238U as fuel. Thus, for example, no allowance is made for fission
product or protactinium losses. The reactors were unreflected, with the
exception of Case 10, which had a 1-ft-thick graphite reflector.

The following objectives were considered in selecting a reference
concept from the cases studied:

1. A breeding ratio near one is desired.

2. The reactor power required should be low (less than 200 Mw).

3. The total uranium inventory should be reasonable.
Table II. Nuclear Characteristics of Several Conceptual MSBE Reactor Configurations

 

 

 

Core Radial Aximl Axial Beactor Primary Peak Power Average Power

c Blank=t Blanket Plena Vessel Salt Systen P’:gi.:nt PM;l:nt 53231 Breeding Rf,q"érfid F;agtion Density Densit;

ase Diameter Helght Salt Thickness Thickness Thickneps® Volume Salt Volumeb F 2363813‘ (xg) Ratio® M (:: r 1) ?_ Cower in Core in Salt

(£t) (£t) Fraction  (ft) (£t) (£t) (£t3) (£¢3) ‘ . & Wi serma ntore  (w/em®) (v/cx®)
1 5 5 0.10 0 -- 0.67 36 100 12 C. 465 159 0.551 108 1.0 99 38
3 3 5 0.10 1.0 - 0.67 92 158 12 0.621 335 0.784 112 0.48 112 25
L 3 5 0.15 1.0 - 0.67 g4 158 12 0.597 322 0.803 108 0.51 116 2L
10¢€ 5 5 0.10 0 - 0.67 36 124 12 G.351 149  0.653 155 1.0 105 Ly
11 3 5 0.15 k.25 -- 0.67 627 724 12 0.616 1524 0.999 17k 0.327 116 9
12 L 5 0.15 4.25 - 0.67 24 825 12 0.467 1316 1.009 183 0.481 112 8
16 L 5 0.15 k.25 - 0.67 124 819 14 0.482  _13k9 1.022 169 0.511 112 7
18 b 6 0.20 3.75 - 0.67 702 Bo6 14 0.493 1357 1.046 188 0.538 112 8
20 b 6 0.20 2.75 - 0.67 60 561 14 0.480 920 1.034 182 0.561 112 11
23 y 6 0.2 2.0 1.0 0.5 379 L8g 14 0.530 886 1.061 199 0.55 102 14
25 i 5 0.2 2.0 1.5 0.5 382 486 14 0.546 906 1.063 188 0.51 103 14
26 % 5 0.2 2.0 1.5 0.5 382 491 12 0.456 765 1.051 158 0.49 105 ik
27 L 5 0.15 2.0 1.5 0.5 378 489 12 0.h22 705 1.0kt 202 0.46 102 15
28 I 5 0.2 2.0 1.07 0.357 337 hap 1k 0.591 B72 1.062 169 0.57 110 1k
29f L 5. 0.2 2.0 1.07 0.357 337 Log 1k 0.492 721 1.069 164 0.59 113 13
30 h 5 0.2 2.0 1.07 0.357 337 43o 12 0.417 616 1.051 169 0.57 112 1k
31 3.5 4.5 0.2 2.0 1.07 0.357 284 362 14 0.587 726 1.0k4g 136 0.52 115 13
33 3 4 0.2 1.75 1.0 0.25 189 255 14 0.723 630 0.993 112 O.hh 118 16
35 3 4 0.2 1.0 1.0 0.25 101 161 14 0.755 415 0.893 100 0.49 119 22
36 3 L 0.2 1.75 1.0 0.25 189 257 12 0.628 551 0.972 115 0.43 118 16
37 4 5 0.2 {Spherical blanket - 8 ft dia) 210 301 12 0.42o L3z 0.994 161 0.59 113 19
38 b 5 0.2 {Spherical blanket - 9 ft dia) 324 418 12 0.418 597 1.039 168 0.57 113 1k
39 L 5 - 0.2 (Spherical blanket - 10 ft dia) k66 565 12 0.413 797 1.075 178 0.55 113 11
4o 3.75 N 0.2 (Spherical blanket - 7.5 £t dia) 182 264 12 0.457 L1z 0.966 143 0.57 114 19
L 3. b5 0.15 (Spherical blanket - 7.5 £t dia) 179 266 12 0.U436 396 0.960 153 0.53 11k 20
MSHE Objective 0.13 12 0.3 ~d < 200 > 0.5 > 100 ~hb

 

8Conteins 6% INOR.

bSYStem volume = vesgsel volume + 0.5 L%i {reactor power in Mw} + 10 £t miscellaneous.
®At start of 1ife with 100% 232U fuel; no absorptions in 138Xe or 233pg,

dReactor power required to achleve a peak damage flux of 5 x 10** neutrons cm™? sec”l.
®Case 10 had a 1 ft thick graphite reflector.

fNew cross sections introduced.

€Tncludes all salt in circulation.
10

L. The concentration of uranium should be near that proposed for
the MSBR, which is 0.3 mole percent.

5. The salt fraction of the core should be near that proposed for
the MSBR, which is 0.13.

6. The average power density of the circulating salt should be
near that proposed for the MSBR, 46 kw/1.

For those cases of most interest, the fast neutron flux In the reactor
vessel wall was determined. The current extent to which specimens of
Hastelloy N have been irradiated is 1 x 102} nvt (En > 0.1 Mev). Until it
is established that the material is adequate beyond this, a concept is pre-
ferred in which the reactor vessel receives a dose of less than this in
its design 1life.

Based on the data in Table II the reactor represented by Case 41 is
judged to most nearly satisfy the requirements of the MSBE. Although the
average power density in the circulating salt is only 20 kw/l, this is
conéidered adequate.

In Case 1 the desired neutron damage flux in the core and a high
average power density in the circulating salt are achieved with a total
power of only 108 Mw. However, the breeding ratio is much lower than
desired and the fast neutron flux in the vessel wall is unacceptably high.
The vessel wall would receive a fluence of 1 x 102! nvt (E, > 0.1 Mev)
in less than 1 year.

In Cases 3, L4, 33, 35, and 36, the fast flux at the vessel wall is
reduced and the breeding ratio is improved by reducing the size of the
graphite core and introducing a salt blanket between the core and the
vessel wall. The desired neutron damage flux in the core is still achieved
with a low total power, less than 115 Mw. However, the 233U concentration
is unacceptably high in all of these cases.

In Case 10 the fast flux at the vessel wall is reduced by & factor
of 10 by introducing a graphite reflector between the core and the vessel
wall. The breeding ratio is too low in this case, however.

The remainder of the cases show the effects of varying the core size,
the volume fraction of the salt in the core, and the blanket thicknesses

in the radial and axial dimensions. Although it 1s possible to achieve a
11

breeding ratio of greater than one, factors such as the total power,
inventory of 233U, concentration of 238U, and average power density in
the circulating salt are either individually, or in some combination,
unfavorable.

Although additional cases will be run before selecting a final con-
figuration of the core, Case 41 is considered to provide a reagonable
preliminary basis for design of the MSEBE.

A concept of the MSBE based on Case 41 is shown in Figs. 1 through 6.
The details indicated in these drawings are schematic only and may be
significantly changed after additicnal analytical studies are made. A
pressure vessel and core configuration is shown in Fig. 1. The reactor
power is 150 Mw(thermal). Primary salt enters the vessel at the bottom
at 1050°F and flows upward through the blanket region between the core
and the vessel wall. About half of the power is generated in the blanket.
Therefore the temperature of the salt rises to 1175°F just before the salt
enters the graphite region. The graphite array is divided into two regions
of equal flow area, the outer region and the central region. After leaving
the blanket region the flow is down through the graphite outer region and
up through the graphite central region at an average velocity of about
10 ft/sec, resulting in a salt temperature of 1300°F at the vessel outlet.

The graphite arrangement in the core is indicated in Fig. 2. The
size of each bar is the same (except for length) as that proposed for the
MSBR. The U4-inch square lattice arrangement in the central region of the
core has a 15% flow area as compared to the 13% in the MBBR. Thies is con-
sidered to be a sufficiently close simulation. The graphite bar at the
cefiter of the core i1s removable through a specimen access port in the
upper head of the vessel. The neutron flux spectrum at this specimen
location is shown in Table III. The peak damage flux (> 50 kev) of
5 x 1014 neutrons/cme— sec, 1s equivalent to an integrated dose of
3 x 10°%2 nvt in two years.

The temperature of this specimen at full power ranges from about
1250°F to 1300°F. The peak gamma heating is about 2.6 w/cm®.

Provisions will be made for inserting surveillance specimens of

Hastelloy N at the upper end of the removable graphite specimen.
ORNL DWG. T0-12350R

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MOLTEN SALT BREEDER EXPERIMENT
REACTOR CORE ASSY.

150 MW T)

 

Figure 1. MSBE Reactor Core Assembly
 

 

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Table III. MSBE Irradiation Facility V-1¥ Neutron Flux

 

 

 

Location Energy Groups
0.9 Mev- 36 Kev- 1.4 Kev- 55 ev- 2.1 ev- 0.8 ev- 0.19 ev- 0.065 ev- 0.007 ev-
1.5 Mev 0.9 Mev 36 Kev 1.4 Kev 55ev 2.1 ev 0.8 ev 0.19 ev -.065 ev
Horizontal (core) 1.8 3.6 3.7 3.3 2.6 0.7 2.1 2.1 0.6
midplane
1 ft above or 1.6 3.1 3.2 2.8 2.3 0.6 2.1 1.8 0.5
below midplane¥**
2 ft above or 1.0 2.0 2.1 1.9 1.4 0.4 1.2 1.0 0.3

below midplane*¥

 

*
L in. square, ~ 4 ft long, vertical at core centerline.

**
Neutron flux, neutrons cm™2 sec™! x 10-'4 (unperturbed).

}__J
=
15

The four control rod locatlons shown in Fig. 2 are also accessible
from above where the rod drives are located. The control rods are cooled
by direct contact with salt flowing upward through the core. With the
control rod drives removed, four additional graphite bars are removable
through the specimen access.

The graphite bars in the outer region of the core are arranged as
shown in Fig. 2. This tongue and groove arrangement is proposed as one
method to separate the flow in the several regions. This permits some
thermal and irradiation expansion or contraction. The graphite array is
held together by hoops at the ends and keyed to prevent rotation. The
entire array i1s supported by a Hastelloy N dish at the bottom of the ves-
sel which in turn is supported by Hastelloy N rods suspended from the ves-
sel upper head. These rods, as shown in Fig; 1, are located about one
foot away from the core at a region where the fast flux (> 0.1 Mev) is
less than 3 x 10'® neut/cm® sec. The fast flux (> 0.1 Mev) at the midplane
of the vessel wall is less than 3 x 10'?, giving the wall an integrated
dose (> 0.1 Mev) of about 1 x 10! nvt in ten full power years. The thermal
flux at the vessel wall is only 2 x 10'° neut/cm® sec.

Using the approach proposed for the MSBR, the entire graphite core
structure 1s replaceable as a unit. A minimum six foot opening is provided
in the top of the vessel to permit removal of the core structure including
the support rods. This vessel opening is extended some 16 ft in height
above the vessel midplane and some 8 ft above the salt level. This permits
location of the vessel closure outside of the cell furnace in a thermally
cool region with a reduced radiation level. Either two concentric metal
gaskets with provision for leak detection between them or a seal weld could
be used in the closure.

The following is proposed for replacement of the core shown in Fig. 1.
After draining the salt and flushing the primary system with inert gas,
the bioclogical shielding blocks are removed, the contaimment seal is broken,
and the pressure vessel seal is broken. Then the upper vessel head, with
the core suspended on the support rods, is hoisted into a carrier. During
this operation the carrier is sealed to the vessel. Large gate valves are
used to isolate the volumes before removing the carrier. The procedure is
reversed in inserting a preassembled replacement unit complete with a new

vessel upper head.
16

Some primary salt will bypass the core during operation. This will
result from leakage between the blanket region and the outlet in the ves-
sel extension. In order to reduce this to an acceptable minimum, graphite
piston rings in grooves on the vessel upper head as shown in Fig. 1 slide
against a raised machined surface inside of the vessel extension just above

the blanket region.

4.3 MSBE Primary Salt Properties

 

The primary salt has the following nominal composition (in mole %):
71.5 "LiF; 16 BeF,; 12 ThF,; 0.5 UF,. (At the start in a system initially
charged with 233U only, the UF, would be 0.4k mole %; at steady state, the
total UF, would be 0.57 mole %). This salt has the physical properties
given in Table I. As indicated in Table I the proposed MSBR salt has a
lower UF, concentration. The difference, however, is small enough to have

little effect on the neutron spectrum, salt properties, and salt chemistry.

L.4 MSBE Primary System

 

The MSBE flowsheet is shown in Fig. 3. This flowsheet is modeled
after that proposed for the MSBR. The flowsheet will be explalned in more

detail under the individual systems.

L.4h.1 Primary Loop

The primary salt leaves a side outlet at the top of the reactor pres-
sure vessel at 1300°F and enters the pump suction. From the pump, the
salt flows to the primary heat exchanger. The flow is down through the
tubes in the vertical heat exchanger, leaving at 1050°F. The primary salt
then enters the bottom of the reactor pressure vessel. For the layout
gstudies, 10-inch pipe is shown between the components except at the pump
suction where 12-inch pipe is used to reduce the velocity entering the
pump.

The total primary salt flow through the core at 150 Mw(thermal) and
with a AT of 250°F is 6.37 x 10° 1b/hr (3850 gpm at 1300°F). The velocity
in the proposed 10-inch pipe is about 16 ft/sec.

The primary system circulating volume exclusive of the expansion

volume in the pump tank is about 266 ft2.
ORNL DWG. T0-11190

    
 

. uny

   
    
        

THTHNA AND WATER  CHARCOAL 8R0S
REMOVAL 90 DAY HOLOUP

PRIMARY GAS RECOVERY SYSTEM

DRAFT STACK

PRIMARY SALY
PRIMARY SALT PROCING DRAIN TANK &
STORAGE TANK

——————

OAS HOLD-UP [ ez vav

     

TRANSFIR SYSTEM

MOLTEN SALT BREEDER EXPERIMENT
SIMPLIFIED FLOW DIAGRAM
150 MW 1

Figure 3. MSBE Flow Diagram

LT
18

The primary piping is of welded construction. Thermal expansion of
the piping is accommodated by allowing the pump and heat exchanger to move
by sliding on their supports. Provisions will be made in the pump inlet
piping for insertion and removal of Hastelloy N specimens in the primary

salt.

L.4.1.1 Primary Pump. The primary system utilizes a single-stage,
sump-type centrifugal pump with an overhung impeller. The pump volute 1is
enclosed in a tank which provides for the system volumetric expansion.

This expansion tank is able to accommodate about 10% of the primary system
salt volume or about 30 ft3. About 89% of the primary salt flow goes

through the reactor vessel, lO% through the gas separation bypass, and the
remainder to miscellaneous components such as the jet pump in the drain

tank. The total flow requirements at a AT of 250°F is L4300 gpm. However,

it is proposed that a larger pump be used in the MSBE to permit operation
with a AT of 200°F at full power, if problems arise at the higher AT. There-
fore a pump with a minimum flow of 5400 gpm is proposed; the head required

is 125 ft. The pump would have a variable speed motor to permit operation
at 4300 gpm.

A single circulating loop is used to reduce the development or extrapo-
lation that will be required in progressing from MSBE size equipment to

equipment for larger reactors.

L.4,1.2 Primary Heat Exchanger. The primary heat exchanger shown in
Fig. 4 has the same type, size (3/8 in. OD) and length (about 21 ft) of

 

tubes as those proposed for the MSBR. The MSBE exchanger, however, will
have only about one-fourth as many tubes as will be in each of the four
heat exchangers in the MSBR. The following conditions are about the same
as those for the MSBR. The velocity of the primary salt in the tubegs is
about 10 ft/sec and the pressure drop about 129 psi. The overall heat
transfer coefficient is estimated to be 950 Btu/hr-£t2:°F when a secondary
salt composed of 92 mole % NaBF, and 8 mole % NaF with an average velocity
of 7.5 ft/sec is used on the shell side.

The tube bundle, as shown in Fig. U4, is removable as a unit using the
maintenance techniques proposed for removing the core graphite array. An

alternate approach to tube repair is to use a tube and shell exchanger,
19

 

 

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150 MW (®)

Figure 4. MSBE Primary Heat Exchanger
20

cut open access hatches to the tube sheets, locate the faulty tube, and

plug both ends.

L.L.2 Gas Separation Bypass

As in the MSBR about 10% of the primary salt flow leaving the pump is
routed through a gas separator. After the primary salt leaves the gas
separator it enters the bubble generator where relatively clean helium
bubbles are injected into the salt. These combined fluids are returned
to the main loop at the pump suction.

The separated gas is combined with overflow salt from the pump ex-

pansion tank and sent to the drain tank.

4.4.3 Primary Salt Drain Tank

 

The drain tank is designed to contain all of the salt from the primary
loop. In addition, it serves as a gas hold-up tank for fission product
decay during operation. A minimum one hour decay of the fission products
in the gas from the pfimary loop is desired before the gas is sent to the
charcoal beds.

The primary loop is drained by gravity to the drain tank after a freeze
valve in the drain line is thawed. The drain valve requires some five
minutes for a rapid thaw. After the valve is open, it takes about twelve
minutes to drain the primary system. The afterheat in the primary system,
twelve minutes after reactor shutdown, is about 2% of the operating power.S
The afterheat in the fission gas from the primary system during operation
ig about 1% of the reactor power.® The primary drain tank heat removal
system is designed for the higher of these two which is 3 Mw.

This heat is removed by a drain tank coolant salt system. This coolant
salt consists of 66 mole percent 7LiF and 3L mole percent BeF,. The coolant
system is composed of ten independent loops, each of which is capable of
removing 12% of the total afterheat. This will permit operation of the
reactor with one coolant loop out of service. Each loop contains a number
of vertical U-tubes in which the coolant salt circulates by natural con-
vection. The heat is removed from the coolant salt by boiling water in a

heat exchanger. The steam is condensed in air cooled coils in a natural
21

draft stack adjacent to the reactor building, and the condensed water is
returned to the water boilers by gravity.

Some primary salt continuously overflows from the pump bowl to the
drain tank and some is entrained in the gas flowing to the drain tank.
This salt is returned to the primary loop by a Jet pump in the drain tank.
This jet is driven with a 40 gpm side stream from the discharge side of
the primary pump.

There is a second jet pump in the drain tank which furnishes salt to
the chemical processing plant. This jet pump is also used to return the
salt to the primary system when it is desired to restart the reactor after

a drain. This jet is driven by an asuxiliary pump of low capacity (around

300 gpm) .

L.4.4 Primary Salt Storage Tank

A primary salt storage tank is provided for the storage of irradiated
salt. In the event of trouble with the primary salt drain tank the salt
can be transferred to the storage tank. The storage tank utilizes a cool-
ing system similar to that of the drain tank but of only one-tenth the
capacity; therefore, the afterheat must be allowed to decay to 0.3 Mw
before the salt is transferred to the storage tank. An alternate approach
for cocling of the storage tank is to use a system similar to that of the
MSRE drain tank in which water was boiled in bayonet-type thimbles in the

drain tank.l©

L.L.5 Primary Salt Sample System

 

It is required that samples of the primary salt be taken while the
reactor is in operation. A sampler-enricher, basically similar to the
one successfully employed in the MSRE, is being considered for the MSBE.
In the MSBE the sample is obtained from the primary salt drain tank since
the salt there is representative of the primary loop and is normally well
mixed by the jet pumps.

In addition, consideration is being given to an on-line analytical
facility which will measure the UF,/UF, ratio and the concentration of

uranium, chromium, and oxygen. The analytical facility is fed by a side
22

stream from the primary loop at the primary heat exchanger discharge.
The primary salt is returned to the primary loop at the suction of the
pump. The use of this system will require development of new instrumen-

tation for these measurements.

4.5 MSBE Secondary System

The composition and physical properties of the proposed MSBR (and
MSBE) secondary salt is given in Table I.

At a power level of 150 Mw(thermal), the secondary salt enters the
shell side of the primary heat exchanger at 850°F and leaves at 1150°F.
The flow rate is 4.7 x 108 1b/hr or 5300 gpm at 1150°F. After leaving
the primary heat exchanger, the salt enters the secondary pump, which is
similar to the primary pump.

Heat is removed from the salt in the steam generator downstream of
the pump. The salt is on the shell side and the steam is in the tubes.
Feedwater is fed to the tubes at TOO°F and 3800 psia. OSteam is generated
with outlet conditions of 1000°F and 3600 psia.

Thermal expansion of the secondary system piping is accommodated by
expansion loops.

Both the secondary side of the primary heat exchanger and the salt
side of the steam generator have relatively high pressure drops and there-
fore require large fractions of the available head of the secondary pump.
A compensating reduction in pressure drop in the secondary system piping
is achieved with a relatively low veloccity of 13 ft/sec, which requires
12-inch pipe.

The secondary salt outlet on the steam generator is a pipe tee with
a rupture disc on one leg. This permits pressure relief in the event of
a tube failure in the steam generator. The line on the downstream side
of this rupture disc is connected to the emergency drain tank which is
also equipped with a rupture disc. Upon rupture of both discs, BF; and
water vapor are released to the cell.

It is desirable to clean corrosion products and water and other con-
tamination from the secondary salt. A small bypass loop arocund the pump
is provided for this purpose. This loop may include a cold trap for corro-
sion products and a nickel tank where batchwise removal of water (possibly

by an HF, BF,, He purge) is accomplished.
23

The pressurizing gas in the pump bowl of the secondary system is con-
tinuously cleaned of tritium, water vapor, radicactive material, and BF,
before being returned to the system as a purge gas for the pump shaft.

The tritium is that which diffuses through the heat exchanger tubing from

the primary system.

4.6 MSBE Steam System

It is required that as much information as practical be obtained
from the MSBE which would be useful in design of the MSBR steam system.
The feedwater conditions (700°F - 3800 psia) and the outlet steam con-
ditions (1000°F - 3600 psia) are therefore the same in both reactors.

The high feedwater temperature is desired to reduce the possibility of
freezing of the secondary salt in the steam generator. The secondary salt
melting point is about T25°F. The high feedwater temperature is achieved
by mixing some of the outlet steam with the feedwater to raise the tempera-
ture above that attainable with conventional feedwater equipment.

The steam generators are fabricated from Hastelloy N, but all other
components in the steam system can be conventional equipment made from
material such as 2-1/4 Cr steel. A water treatment plant is available
for control of the water chemistry. Provisions are made to evaluate the
metallurgical, corrosion, and deposition problems in the steam system.

A steam turbine is not considered necessary in the MSBE in obtaining
information for the MSBR but could be used if desired as indicated by

economic studies.

4.7 MSBE Reactor Cell

The reactor components must be maintained above the freezing point
of the primary salt even during zero power operation. This is achieved
by using the reactor cell as a furnace. Heater thimbles penetrate the
top of the cell and are capable of maintaining the interior at 1000°F.
The reactor cell is shown in Figs. 5 and 6. The cell liner is about
26 ft in diameter and 34 ft in height. The cell contains a nitrogen

atmosphere.
2k

ORNL DWG. T0-12349R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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POLAR CRANE -
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SECONDARY
PRIMARY. PUMP
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MOLTEN SALT BREEDER EXPERIMENT- -
REACTOR BUILDING SECTION B-B B%i‘f,&‘?fims‘*”

Figure 5.

SO MW (1)

MSBE Reactor Building Section B-B
 

 

 

 

 

 

 

 

 

 

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" FUTURE CHEMIC:AL ORNL DWG. T0-12351
.' PROCESSING BUILDING X

S T g S T ——— —— e S R S——

PRIMARY SALT
STORAGE TANK

PROCESSING HEAT EXCHANGER

{

|

|

|

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SECONDARY SALT
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REACTOR
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MOLTEN SALT BREEDER EXPERIMENT
REACTOR BUILDING PLAN A-A
IS0 MW (v)

figure 6.

MSBE Reactor Building Plan A-A

a2
26

Thermal insulation with an inner liner of stainless steel is used
inside of the cell liner to reduce the heat losses. Any salt which legks
onto the stainless steel liner remains molten and is routed through a
special line to the primary system drain tank. An air-cocled-steel thermal-
shield surrounds the cell liner to prevent radiation overheating of the
concrete biological shielding.

Removable shielding plugs and flanged nozzles on the cell liner per-

mit access to the components for maintenance.

4.8 Drain Tank Cell, Off-Gas Cell, and Secondary Cell

The primary system drain tank cell is a furnace similar to the reactor
cell. The heater thimbles are in the annular region between the drain
tank and the cell liner. This cell liner which will catch any salt spilied
from the drain tank, is cooled by thermal radiation to the cell walls.

The primary system off-gas cell is a shielded contaimment cell. Since
there is no molten salt in this system, there are no heating requirements.

The secondary cell (or steam cell) contains the secondary system com-
ponents including the steam generator. These components must be maintained
at a temperature above the freezing point of the secondary salt. This is
achieved with the furnace concept for the cell as used in the reactor cell.

The secondary salt in the primary heat exchanger will incur some
induced activity. The secondary system is therefore in a shielded con-
tainment cell. This cell is designed to withstand the pressure which
would result from a ruptured steam tube. Two block valves on each steam
line and feedwater line are used to minimize the pressure in the cell as

a result of such a rupture.

4.9 MSBE Reactor Building

 

An elevation of the reactor building is shown in Fig. 5, and a plan
of the cells within the building is shown in Fig. 6. The building serves
as the containment during maintenance through the shielding plugs above
the cells. The building is cylindrical with a hemispherical dome for a
roof. A polar crane, located at the top of the cylindrical portion of

the building, is used for movement of equipment within the building.
27

4.10 MSBE Chemical Processing

The chemical processing plant is similar to that proposed for the
MSBR. The equipment is located in the chemical processing building adja-
cent to the reactor building. The small lines connecting the two plants
are equipped with isolation valves. This permits installation, alteration,
and maintenance of the processing eguipment while the reactor is in
operation.

The chemical processing flowsheet proposed for the M3BR!! is given
in Fig. 7. The processes are described in more detail elsewhere;ll g
brief description 1s given below.

Salt withdrawn from the reactor is fed to a fluorinator, where most
of the uranium is removed as UFg. Most of the salt leaving the fluorinator
is fed to a reductive extraction column, where the remaining uranium is
removed and the protactinium is extracted into a bismuth stream. The
bismuth stream containing the extracted protactinium flows through a tank
of sufficient veolume to contain most of the protactinium in the reactor
system. Most of the bismuth stream leaving the extraction column is con-
tacted with an H,-HF mixture in the presence of about 10% of the salt
leaving the fluorinator in order to transfer materials such as uranium
and protactinium to the salt stream. This salt stream is then recycled
to the fluorinator.

The bismuth stream leaving the lower column also contains several
materials that must be removed for satisfactory operation of an MSEHR.

The most important of these are fission product zirconium, which can be
an important neutron absorber, and corrosicn product nickel, which forms
an intermetallic nickel-thorium compound having a low solubility in bis-
muth. These materials and others that do not form volatile fluorides
during fluorination are removed by hydrofluorination, in the presence of
a salt stream, of a small fraction of the bismuth stream exiting from
the lower column. The salt is then fluorinated for removal of uranium.
Sufficient time is allowed for the decay of 233Pg so that the rate at
which this material is lost is acceptably low. The remaining materials,
including Zr, Ni, 281Pg, and Pu, are withdrawn in the salt stream from

the fluorinator. Oxidation of part of the metal stream leaving the lower
 

 

 

 

ORNL- DWG 70~-2833

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SALT - UFg PROCESSED SALT
PURIFICATION REDUCTION - ':
a f |
)
[ H EXTRACTOR !
| 2 ! 8i
| UFg I
| _ SALT CONTAINING RARE EARTHS '
i e ——————————— - ————————— 9 f - - ——
UFg | [
COLLECTION : :
f l’“"‘ EXTRACTOR :
: EXTRACTOR :
l J
' - BI"LI
Y | L
UF, - Po LiCt
REACTOR : 6 ~ —— DECAY ! FRACTION Li)
I EXTRACTOR
i
: EXTRACTOR UFg _l-rr‘
i T L— - &Bj—
+ DIVALENT
' - SALT ™ ™7  RARE EARTHS
FLUORINATOR - o OXIDIZER[ FLUORINATOR = TO |
: , SALT WASTE j — Son B‘il-oLiE
0.05 MOL
? i : f f EXTRACTOR : FRACTION Li)
F H, —HF F
2 | 12 2 f— — — > Bi-Li
- O LR
N
oxiDIZER |- L — — — - ol REDUCTANT | _ o T2
SALT CONTAINING Bi
PaF, AND UFg4 ' T
Ho-HF Li, Th

Figure 7.

MSBE Chemical Processing Flowsheet

____.._.._____.,.________.___.._1

-

8c
29

contactor is chosen as a means for removal of these materials, since this
results in discard of no beryllium and very little lithium or thorium;
discard of salt from other points in the system would result in much higher
removal rates for the major components LiF, BeF,, and ThF, .

The bismuth streams leaving the hydrofluorinators are then combined,
and sufficient reductant (iithium and thorium) is added for operation of
the protactinium isolation system. Effectively, this stream is fed to
the upper column of the protactinium isolation system; actually, it first
passes through a captive bismuth phase in the rare-earth removal system
in order to purge uranium and protactinium from this captive volume.

The salt stream leaving the upper column of the protactinium isclation
system contains negligible amounts of uranium and protactinium but contains
the rare earths at essentially the reactor concentration. This stream is
fed to the rare-earth removal system, where fractions of the rare earths
are removed from the fuel carrier salt by countercurrent contact with
bismuth containing lithium and thorium. The bismuth stream 1s then con-
tacted with LiCl, to which the rare earths, along with a negligible amount
of thorium, are transferred. The rare earths are then removed from the
LiCl by contact with bismuth containing a high concentration of 7Li.
Separate contactors are used for removal of the divalent and trivalent
rare earths in order to minimize the quantity of 7Li required. Only
about 2% of the LiCl is fed to the contactor in which the divalent mate-
rials are removed.

Small scale testing of these processes 1is now 1in progress. Material
for the equipment is being developed. The plant proposed for the MSBE is
of such size that determination of most of the problems in an MSBER size

processing plant is possible.

5. MSBE EXPECTED ACCOMPLISHMENTS

The design, procurement, installation, and maintenance of equipment
for the MSBE will give essential information for design of similar equip-
ment for larger reactors. The operation of the MSBE will demonstrate the
reliability of equipment and systems from performance and safety stand-

points. The evaluation of material samples and radioactive deposits will
30

give important information on material behavior and fission product be-
havior in an MSBR environment. The central core graphite specimen can be
removed and replaced individually after several months operation, and it
is expected to replace all of the core graphite after two years at fuil
power.

The MSBE will be an important step in continuing the training of
operating personnel, the development of MSBR preliminary operational
procedures, and the development of safety criteria for the MSBR.

The MSBE will be used in conjunction with its processing plant to
study the nuclear characteristics of a reactor with a breeding ratio near
one. The irradiated salt available to the processing plant will permit
development work on a processing plant at conditions very close to those
of an MSBR.

A six year program of operation of the MSBE at full power should be
sufficient time to achieve the initial objectives. Continued use may be
important for irradiation of improved graphite, testing of improved designs
of equipment such as steam generators, improvement of chemical processes,

and the study of the long term effects of operation on salts and materials.

6. APPENDIX I

THE MOLTEN SALT REACTOR EXPERIMENT

6.1 Description

The MSRE was a reactor in which a molten fluoride salt containing
fissile material (but no thorium) was circulated through a single-region
core of bare graphite bars. All metal in contact with the molten salts
was standard Hastelloy N. The flow diagram is shown in Fig. 8; some
additional operating data were given previously in this report in Table I.
The design, development, and construction of the MSRE are described in
Ref. 3. Ref. 10 is a very detailed description of the reactor design.

The brief description which follows is intended to focus on the features
that are of particular interest in a consideration of the bases for future

molten salt reactors.
 

 

 

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1 1 E-.L__ WATER  STEAM !_fl FILTERS
! | WATER STEAM . ; .
4 MAIN i t i
I CHARCOAL U : 1
BED .
= CHARCOAL :
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i | fuel i
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[ L TanK NO. 1 |
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S0DIUM
FLUORIDE BED

 

Figure 8. MSRE Flowsheet

T¢
32

The 54-in.-diameter MSRE core was comprised of bare graphite bars,
2-in. square and 6L4-in. long. The graphite bars had a channel machined
in each face to form 1.2 x O.4 in. flow passages between bars. The graph-
ite (Grade CGB) was especially produced to limit pore size to 0.4 microns
to keep out the salt (which does not wet the graphite). Fuel salt passed
upward through the core channels in laminar flow. A removable, 2-in.-diam-
eter, Hastelloy N basket held specimens of metal and graphite in a special
channel near the center of the core. Three thimbles (also near the core
centerline) housed flexible control rods used for temperature regulation
and shutdown.

The fuel salt was circulated at 1200 gpm through the shell side of
a U-tube heat exchanger, where heat was transferred to a secondary salt
(LiF-BeF,, 66 — 34k mole %). The coolant salt, circulating at 850 gpm,
delivered the heat to an air-cooled heat exchanger. The heat transfer
capability of this system limited the power that could be dissipated to
7.3 Mw when the core operated at its normal temperature of 1210°F.

The salt pumps were centrifugal pumps of conventional hydraulic
design with vertical, overhung shafts having ocil-lubricated bearings.

The volute of each pump was submerged in a pool of salt in a tank (pump
bowl) that contained the surge space for the loop. The pressure in the
blanket gas (and at the open pump suction) was 5 psig. Pump discharge
pressures were 55 psig in the fuel loop and 70 psig in the coolant loop,
respectively. A tube into the top of the fuel pump bowl connected to the
sampler-enricher, contained a moctor-driven reel by which sample buckets
or capsules of enriching salt (LiF—UF4, T3 — 27 mole %) could be lowered
into the salt pool. A similar device was provided on the ccolant pump.

A spray ring in the top of the fuel pump bowl took about 50 gpm of
the pump discharge and sprayed it through the gas above the salt. The
purpose was to provide contact so that the gaseous fission products could
escape into the gas. A flow of 3 liters/min of helium carried the xenon
and krypton out of the pump bowl, through a holdup volume, a filter
station, and a pressure-control valve to charccal beds. These consisted
of pipes filled with charcoal, submerged in a water-filled pit at about

90°F. The beds, operated on a continuous-flow basis, delayed xenon for
33

about ninety dasys and krypton for about seven days so only stable or long-
1lived nuclides got through to the stack.

All salt piping and vessels were electrically heated to prepare for
salt filling and to keep the salt molten when there was no nuclear power.
Heaters in the reactor cell were incorporated in removable, reflective-
metal insulated units. Thermocouples under each heater monitored tempera-
tures to avoid overheating the empty pipe. (The salt pumps were used to
circulate gas during system heatup and cooldown to help maintain uniform
temperatures.)

There were no mechanical valves in the salt piping. Flow in drain
and transfer lines was blocked by plugs of salt frozen in flattened sec-
tions of the lines. Temperaturesin the freeze valves in the fuel and
coolant drain lines were controlled so they would thaw in 10 to 1> minutes
when a drain was requested. A power failure of longer duration also re-
sulted in a drain because thé cooling air required to keep the valves
frozen was interrupted. The drain tanks were almost as large as the
reactor vessel, but the molten fuel when in the drain tanks and away from
the graphite was subcritical. Water-cooled bayonet tubes extended down
into thimbles in the drain tanks to remove up to 100 kw of heat, if neces-
sary. Steam produced in the tubes was condensed and returned by gravity.

The physical arrangement of the equipment is shown in Fig. 9. The
reactor and drain tank cells were connected by a large duct so they formed
a single containment vessel. The tops of the two cells consisted of two
layers of concrete blocks, with a weld-sealed stainless-steel sheet be-
tween the layers, and the top layer fastened down. A water-cooled shield
around the reactor vessel absorbed most of the escaping neutron and gamma-
ray energy. The cell atmosphere was kept at 140°F by water-cooled, forced-
air space coolers., The cooling air for the freeze valves and the fuel-
pump bowl was actually reactor-cell atmosphere, compressed and cooled. A
fraction of the blower output was discharged pest & radiation monitor and
up the ventilation stack to keep the reactor and drain tank cells at
-2 psig during operation. A small bleed of nitrogen into the cell kept
the oxygen content at 3% to preclude fire if fuel-pump lubricating oil

spilled on hot surfaces.
     

REACTOR CONTROL
ROOM

  

Figure 9.

3L

> REMOTE MAINTENANGE
“ CONTROL ROOM {

 

 

. REACTOR VESSEL
HEAT EXCHANGER
. FUEL PUMP
FREEZE FLANGE
THERMAL SHIELD
. COCLANT PUMP

s N

Layout of the MSRE

ORNL-DWG 63-1209R

|
|

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RADIATOR

. COCLANT DRAIN TANK

. FANS
. FUEL DRAIN TANKS

FLUSH TANK

. CONTAINMENT VESSEL
. FREEZE VALVE
35

The 5-in. salt piping in the reactor cell included flanges that
would permit removal of the fuel pump or the heat exchanger. These
flanges were designed to form a frozen salt seal in the gap between the
flange faces. This seal was backed-up by a metal ring joint seal to
prevent escape of gaseous fTission products. The uninsulated flanges
were maintained below the melting point of the salt by thermal radiation
to the reactor cell environment.

All the components in the reactor and drain tank cells were designed
and arranged such that they could be removed by the use of long-handled
tools from above. When maintenance was done, the fuel was secured in a
drain tank and the connecting lines were frozen. The upper layer of blocks
was removed, a hole was cut in the membrane over the item to be worked on,
and, after a steel work shield was set in place, a lower block was removed.
A large duct from the reactor cell to the upstream side of the ventilation
filters was opened to draw air down through the shield openings. Tools,
lights, and viewing devices were inserted through fitted openings in the
work shield and items to be replaced were unbolted so they could be lifted
with the building crane.

In the same building adjacent to the drain tank cell, there was a
simple facility for processing the fuel or flush salt. The purpose was
twofold: +to remove oxide contamination from the salt if this should be
necessary, and to recover the uranium from the salt. One whole batch of
salt (about 75 ft8) could be transferred into the tank where it was sparged
with hydrogen fluoride gas to remove H 0 and convert the metel oxides to
fluorides. A fluorine gas sparge was used to convert UF, to the volatile

UFg . The UFg gas was then trapped on a bed of sodium fluoride pellets.

6.2 Experience

Design of the MSRE was started in the summer of 1960. The design,
component development, procurement, and construction were conducted be-
tween this time and October 1964 when molten salt was first loaded in
the MSRE. In comparison with other reactor experiments, the MSRE ran
long and well. At the time of the final shutdown in December 1969, salt
had circulated in the fuel system for 21,788 hours, the reactor had been
36

critical for 17,655 hours and had produced heat equivalent to 13,172 full-
power hours. During this ample period of operation, the MSRE produced
much new and valuable information and accomplished its goal of demonstra-
ting the practicality of the molten salt reactor concept. Experience in
the most important areas is summarized below. (A convenient review as

of July 1969, appears in Ref. 12.)

6.2.1 Fuel Chemistry

The experience with regard to the chemical stability of the MSRE
fuel salt was excellent. It had been anticipated that the fuel might be
subject to uranium separation under two extreme conditions, but neither
was approached during MSRE operation.

Oxide contamirnation, which could lead to UOQ, precipitation, was con-
trolled by providing a blanket gas (ordinarily helium) from which oxygen
and moisture had been removed by passage through a 1200°F titanium sponge.
On occasions when the reactor vessel was open, moist air intrusion was
minimized by first filling the system with dense argon and working through
a nitrogen-purged standpipe. Fuel salt analyses showed that the oxide
level remained remarkably low (~ 60 ppm), well below any solubility limit.
As a result of this experience, ZrF,, which was in the MSRE fuel as a
buffer against U0, formation, is not considered necessary in fuels for
future molten salt reactors.

The other mechanism that was considered credible at the time the
MSRE was designed was reduction of uranium from the fluoride to the metal-
lic state as a result of the fission process. There was some uncertainty
in the valence state that several fission products would assume in the
MSRE fuel salt environment, and it was recognized that if the average
products of one fission reacted with more than 4 fluorine atoms, some UF,
would be reduced to UF; as a result of power operation. If this were un-
checked, the UF, concentration conceivably could get high enough to dis-
proporticnate into UF, and insoluble U. It turned out that the reverse
was true: the fission products tied up slightly less than 4 fluorine
atoms on' the average, so Ut was very gradually raised to U+t. To offset

this tendency and keep the system slightly reducing (to avoid Hastelloy N
37

corrosion), it was necessary only to make small (~ 10-g) additions of
beryllium metal through the sampler-enricher at intervals of a month or

two .

Coulometric analysis of fuel salt samples for uranium showed good
reproducibility and high precision (+ 0.5%) and the indicated inventory
changes agreed well with calculated burnup. A far more sensitive (al-
though less unequivocal) indication of uranium concentration was the
reactivity balance by the on-line computer. This never indicated any
anomalous behavior of the uranium.

Eerly operation with partially enriched uranium built up about 0.6 kg
of plutonium as PuF; in the MSRE fuel. The stable behavior of plutonium
in the MSRE and laboratory verification of adequately high solubilities
of PuF, and Pu,O, made it appear reasonable to fuel the MSRE entirely
with plutonium. This was not done, but during 1969, capsules of PuF,
were added to the fuel salt to compensate for fissile material burnup.

The behavior of fission products was intensively investigated in the
MSRE. The noble gases were stripped into the offgas as expected and most
of the other fission products stayed in the salt, also as expected. The
exception was the "noble-metal” group - Mo, Nb, Ru, and Te, whose probable
behavior could not be predicted with certainty before the MSRE operatead.
It appeared that they existed in the normal reducing environment of the
MSRE in elemental form, as colloidal particles, which tended to go to the
metal and graphite surfaces and to accumulate at the gas-liquid interfaces.
Generally only a few percent of the inventories of these elements were
found in the salt. Gas samples indicated that a few percent were carried
into the offgas system, implying that the bulk was on the metal and graph-
ite. This was supported by radiocanalysis of core specimens and measure-
ments of deposition in the heat exchanger by remote gamma-ray spectrometry.
This information is quite important, for it means that high-power molten
salt reactors must deal with the afterheat problems resulting from depo-
sition of the noble metals.

Tritium was produced in the MSRE fuel salt at a calculated rate of
4O Ci/day (35 Ci/day from €Li and 5 Ci/day from 7Li). Measurements showed
that the amount in the fuel offgas leaving the charcoal beds rose gradually
38

to about 25 Ci/day during sustained power operation. 3Both this obser-
vation and the determination of tritium in materials exposed briefly in
the fuel pump implied significant internal holdup and gradual release of
tritium. Some tritium diffused through the heat exchanger tubes in the
MSRE: 0.6 Ci/day was found in the coolant salt offgas and around 3 to 5
Ci/day appeared in the air going up the ccolant stack. These rates were
lower than predicted by accepted relations for tritium diffusion and

suggest the need for refinements in the calculations.

6.2.2 Materials

Compatibility. The compatibility of the salt, the graphite moderator,
and the Hastelloy N container material was demonstrated by analysis of
hundreds of samples of fuel salt and examination of specimens exposed to
salt in the core for as long as 14,714 hours.

The corrosion of the Hastelloy N in the fuel system was very slight,
consisting primarily of leaching of chromium from the surface. (For a
short time after the fuel processing and 223U loading there was apparently
some oxidation of iron also.) Analyses of the fuel salt showed a chromium
increase over the 48 months from December 1965 to December 1969, equivalent
to all the chromium in a 0.46-mil layer over the circulating loop surfaces.
Metallographic and microprobe examination of core specimens showed no void
formation due to leaching, but there was a chromium concentration gradient
from the surface to a depth of 2 to 3 mils. Carburization of metal speci-
mens in contact with graphite extended to a depth of only 1 mil.

In the coolant salt system there appears to have been virtually no
corrosion. The chromium concentration in the coolant salt remained at
37 £ 7 ppm through 26,076 hr of circulation and preliminary examination
of radiator tube specimens shows no evidence of deposition by mass transport.

Graphite specimens exposed to fuel salt in the core for periods up
to 14,714 hr showed no attack; there were no changes in surface finish
and no cracks other than those present before exposure. Only extremely
small quantities of salt were found to have penetrated the graphite either
through pores or cracks. (Assuming the specimens are typical, there is

less than 10 g of uranium in the 3700 kg of graphite in the core.)
39

Irradiation Effects. Specimens of the standard Hastelloy N heats

 

used in constructing the MSRE, after exposure to thermal neutron fluences
up to 1.5 x 10°! n/cmz, showed drastically reduced fracture strain and
creep rupture life at high temperatures. (Fracture strains at 650°C were
less than l% for some specimens.) Although these effects did not become
limiting in the MSRE, they would make the standard alloy unsuited for
high-flux, long-term applications. Specimens of Ti- and Zr-modified
Hastelloy N exposed in the MSRE core proved to have excellent corrosion
resistance and far less decreagsge in creep rupture life and fracture strain
due to irradiation than the standard alloy.

Graphite specimens showed no measurable irradiation effects at the

fluences attained in the MSRE.

6.2.3 Nuclear

Confidence in the predictions of nuclear performance of future molten
salt reactors is bolstered by results of measurements in the MSRE. Agree-
ment of calculated and observed critical concentrations of fissile material
was good, both in the 235U and the 233U startups. Changes in the isotopic
ratios of uranium in the MSRE fuel and encapsulated uranium in the MSRE
core were measured very precisely. From these measurements most accurate
values were obtained for neutron yields (n) for 238U and 238U in a neutron
spectrum gquite similar to that in molten salt breeder reactors.

The dynamic behavior of the system agreed with calculations which
took into account effects of transport of delayed neutron precursors in
the circulating fuel. As predicted, the stability margin of the MSRE

was quite satisfactory.

6.2.4 Equipment

Conservative design and great care in quality assurance (plus the
chemical stability and compatibility of MSRE materials) were responsible
for the high degree of reliability that was attained in the MSRE. This
reliability was demonstrated over the last fifteen months of 238U operation
when the reactor was available 87% of the time {critical 80% and involved

in the removal of core specimens T%).
Lo

Salt Pumps. The two original salt pumps served throughout the MSRE
operation, accumulating 21,788 hr of salt circulation on the fuel pump
and 26,076 hr on the coolant pump. The only problem with the pumps was
0il leakage from the shaft seal drainage passages into the pump bowls.

The oil, which leaked at roughly 5 ce/day into each pump, had no delete-
rious effect on the salt, but its thermal decomposition products accumu-
lated at points in the offgas system. The spare pump rotary elements
were seal-welded to prevent such leakage, but since the offgas problems
were adequately handled with the original pump rotary elements, the spare
pump rotary elements were not installed.

Offgas System. The fuel offgas system caused some delays during the
initial power ascension when oil that had collected in the lines during
the prenuclear testing was vaporized and polymerized by the gaseous fission
products. This was overcome by installation of larger valves and a 3-stage
filter. 0il in the coolant offgas system gradually accumulated and even-
tually required cleanout of the lines near the pump bowl. It was neces-
sary to rod out the fuel offgas line at its outlet from the pump bowl due
to accumulation of frozen salt droplets (originating in the xenon stripper
spray in the pump bowl). After the third such operation in 24 months, a
heater was installed so salt accumulations could be melted ocut. Holdup
of xenon and krypton on the room-temperature charcoal beds met design pre-
dictions and showed no evidence of decrease in capacity.

Heat Exchangers. Measurements on the MSRE heat exchangers confirmed

 

that conventional design calculations adequately predict molten salt heat
transfer and showed no change in performance over the many months of power
operation. The original underestimates of performance resulted from the
use of incorrect values for salt thermal conductivity and iImproper calcu-

lation of the air-side coefficient in the air-cooled heat-exchanger.

6.2.5 Maintenance of Radioactive Systems

 

Practical maintenance of highly radioactive systems was demonstrated
in the MSRE. The basic philosophy was one which can be used, with appro-
priate implementation, in large, high-power molten salt reactors. With
careful design to permit use of simple tools, operated remotely or semi-

remotely, the MSRE approach can be used to replace virtually any component.
Ly

Although there was little trouble with the major components in the MSRE,
enough Jjobs were done to thoroughly test this scheme. The semiremote
maintenance took longer than equivalent non-radioactive Jjobs, but could
be scheduled with confidence. By the use of temporary containment enclo-
sures and shielding, radiocactive contamination was controlled and person-

nel exposures to radiation were held well below permissible limits.

T. APPENDIX TIT

7.1 MSBR Plant Description (5,6)

 

A simplified flow diagram of the primary and secondary systems is
shown in Fig. 10. The salt composition and some of the design conditions
are given in Tsble I in the main body of this report. The thermal rating
of the plant is 2250 Mw. The primary salt is circulated by four pumps
operating in parallel and is heated from 1050°F to 1300°F in its passage
through the core. Each pump has a capacity of about 16,000 gpm and cir-
culates the salt through one of four primary heat exchangers and returns
it to a common plenum at the bottom of the reactor vessel.

There are four corresponding secondary loops. Each of the four
secondary salt circuits has a pump of 20,000 gpm capacity which circulates
the secondary salt from the corresponding primary heat exchanger through
steam generators in adjacent cells. The secondary salt is heated from
850°F to 1150°F in a primary heat exchanger. The steam generated is at
1000°F and 3600 psia.

A plan of the reactor plant is shown in Fig. 11 and a sectional ele-
vation in Figs. 12 and 13.

The reactor cell houses the reactor vessel and the four primary heat
exchanger~circulating pump loops. There are removable plugs over all
components which might require maintenance. The reactor cell is normally
maintained at a temperature of about 1000°F by electric heater thimbles
to insure that the salts will be molten. This "furnace" concept for heat-
ing is preferred over trace-heating of lines and equipment because it
should give more uniform heating, heater elements can be replaced more
easily, and there is no need of thermal insulation on equipment that would

crowd the cell and require removal for maintenance and inspection.
 

He SUPPLY

   
  

 

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CHEMICAL PROCESS PURGE
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REACTOR CELL SECONDARY EXHAUST
MAINTENANCE VENT LINE

   
   
     
   
 
    
 
 
 
     
 
 

IODINE TRAP
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PRIMARY SALT
STORAGE TANK

Figure 10. TFlow

 

 

PRIMARY SALT FREEZE VALVE
DRAIN TANK AND
W= R Gas HOLD uP

FLOW RATES ARE TOTAL FOR
1000 Mw(e) PLANT

Diagram for MSBR Plant
192 ft Oin.

ORNL.- DWG 69-1049tA

 

= <o e 230ft Oin. - : -
48ft Oin, 30ft Qin. =l - 30 Qin, 30t Oin. 30 ft Oin ~ 48t Oin.

 

   
 

      
 

             
    

   

   

STEAM CELLS

  
  
   
 
 

 
 

46ft Oin. NO. 2 NO. 3

   

 
 
  

REACTOR CORE
ASSEMBLY AREA

 
  
  
 

REACTOR COMPLEX
CONTAINMENT

{134ft Cin.1D)

 
  
 
 
   
  

REACTOR CELL

INSTRUMENTA-
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128 O

      

87 ft Oin.

   
   

. NEW CORE
AUXILIARY EQUIPMENT REPLACEMENT CELL AIR LOCK

   

    
  
  
   
 
   

 
 
  
  

 
  

CHEMICAL

: FOR

1 CHEMICAL PROCESSING

: PROCESSING
CELL

ACCESS TO
FREEZE VALVE CELL
WORK AREA

   

CONTROL ROD
STORAGE
CELL

OFF-GAS CELL

   
    

ORAIN TANK AND
OF F-GAS HEAT RE-
MOVAL SYSTEM
AREA

CHEMICAL PROCESSING
HEAT REMOVAL
SYSTEM AREA

   
  
    
 
  
  
  
   

 
 
    
 

351t Qin.

/

    
 

SPENT CORE
STORAGE CELL

SPENT MEAT EXCHANGER
STORAGE CELL

FUEL STORAGE

TANK CELL \

Figure 11. Plan View of MSBR at Reactor Cell Elevation

 
 

 

 

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ORNL-DWG 69-10489A

     
 
       

    

HIGH BAY
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CRANE BAY
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UPPER LEVEL
—_—

 

 

 
 
 
 
 

REACTOR CELL

 

STEAM CELL

    

 

 

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LOWER LEVEL HOT CELLS

   

COOLANT SALT
DRAIN TANK CELL

  

WASTE STORAGE CELL

 
   

 

 

 

 

   
 
 
 

 

 

 

 

AUXILIARY
EQUIPMENT CELL

o

 

 

 

  

 

 

 

 

 

 

 

Figure 12. Sectional Elevation Through MSBR Plant Building

T
  
 
   
 
   
   
  
  
   
    

AIR COOLED RADIATORS IN STACK

STEAM

  

 

DRAIN TANK COOLANT SALT
(LBz, 40 CIRCUITS)

      
  

 

   
    
  
     

DRAIN TANK AND
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-

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PRIMARY
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1

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DRAIN LINE
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Figure 13.

i
7 7

Elevation of MSBR Drain Tank Cell

1
L6

The stainless steel "catch pan' at the bottom of the reactor cell,
shown in Fig. 13, slopes to a drain line leading to the primary salt drain
tank located in an adjacent cell. 1In the very unlikely event of a major
salt spill, the salt would flow to the tank. A fusible valve is provided
in the drain line to isolate the tank contents from the cell during nor-
mal conditions.

High temperature thermal insulation is attached to the inside of the
cylindrical wall and the upper and lower heads of the reactor cell con-
taimment vessel to limit heat losses from the reactor cell. The inside
surface of the insulation is covered with a thin stainless steel liner
to protect the insulation from damage, to act as a radiant heat reflector,
and to provide a clean, smooth surface for the interior.

The primary drain tank cell is essentially the same "furnace" and
containment concept as the reactor cell.

The four steam-generating cells are located adjacent to the reactor
cell. These house the secondary-salt circulating pumps, the steam genera-
tors and the reheaters. The cell construction is similar to that of the
reactor cell but only a single contaimment is used. These cells are also
based on the "furnace" concept.

Through careful quality control of materials and workmanship, the
reactor loops containing fission products would have a high degree of
integrity and reliability in preventing escape of radicactive materials
from the system.

The MSBR core design is based on replacement of the core graphite
after an integrated neutron dose of about 3 x 102 neutrons/cm? (for
E > 50 kev). At a peak core power density of about 65 kw/liter, the
enticipated core graphite life is about U4 years. The breeding perfor-
mance at this power density is considered adequate. The calculated
doubling time is 21 years. The dimensions of the reactor were obtained
through nuclear physics optimization studies discussed in Ref. 5.

A plan and elevation of the reactor vessel assembly are given in
Figs. 14 and 15. The vessel is about 22 ft in diameter containing a
graphite structure some 20 ft high. For the proposed 7O psi inlet pres-
sure, the Hastelloy N vessel would require an estimated 2 in. cylindrical

thickness.
GRAPHITE
REFLECTOR--._

REACTOR
VESSEL

b7

_SALT TO PUMP -
(taTotaL) e

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Figure 14,

Plan View of MSBR Vessel

ORNL- DWG 69-6803

.
L8

ORNL-DWG 69-6003

   
   
   
   

   

e — 48 ft Qlin,

 

 

 

 

REACTOR
COVER — — -
i
REACTOR
VESSEL-—————___
I
|
CONTROL RODS — l
GRAPHITE
REFLECTOR ‘
SALT TO PUMP
(4 TOTAL)7
S
.
i
GRAPHITE 1
REFLECTOR '
13t Oin,
320 10,
MODERATOR
ELEMENTS —- |
REFLECTOR
RE TAINING RINGS
i
CLEARANC ;
SPACE - ——— " P
‘ ZONE TI
!
GRAPHITE o 6 ft 6in,
REFLECTOR - — 4‘
SALT FROM HEAT |
EXCHANGER (4 TOTAL) —
"\— 4 —— 1

DRAIN e

Figure 15. Sectional Elevation of MSBR Vessel
49

A cylindrical extension of the vessel above the salt overflow level
in the pump tank permits location of the vessel closure above the roof
of the reactor cell "furnace." This makes possible a joint which can be
remade after replacement of the core graphite.

The fuel salt enters the bottom of the reactor vessel at 1050°F,
flows upward and leaves at the top at about 1300°F. The core Zone 1 is
about 14 ft in diameter and consists of extruded graphite elements, 4 in.
square x 13 £t long. A minimum 0.6 in. diameter hole through the center
of each element, and ridges oriented to separate the pieces, furnish flow
passages and provide the requisite 13% salt volume in this most active
portion of the reactor.

Special shaped extensions on each end of the elements provide a
greater salt-to-graphite volume ratio at the top and bottom of core Zone 1
to form an undermoderated region which helps reduce the axial neutron
leakage from the core. By varying the salt velocity, a uniform temperature
rise across the core is obtained. The maximum salt velocity in the core
is about 8 ft/sec. The overall pressure drop in the salt flowing through
the core is about 26 psi.

Core Zone 2 consists of graphite slabs 2 in. thick and 13 ft long
arranged radially around core Zone 1 to form a 17 ft diameter region.

The salt volume in core Zone 2 is about 37%. This under-moderated region
serves to reduce the radial neutron leakage. The slabs provide the stiff-
nesg to hold the inner core graphite elements in a compact array as dimen-
sional changes occur in the graphite. The slabs are confined by graphite
rings at the top and bottom. Graphite in the central region of core

Zone 1 would require replacement after four years. It was decided to
replace all of the graphite in core Zones 1 and 2 as a unit. This permits
the replacement core to be assembled as a unit in a clean area.

Surrounding core Zone 2 in the radial direction is a 2 1/2 ft thick
graphite reflector. This graphite receives a relatively low neutron dose
and is structurally suitable for the 1life of the reactor. The reflector
is composed of wedged-shaped slabs, sbout 1 ft wide at the thicker end
and about 4 ft high. These slabs are spaced about 1/L4 in. from the ves-
sel wall to allow an upward flow of fuel salt to cool the metal wall and

maintain it below the design temperature of about 1300°F. The reflector
50

has a radial clearance of about 1 1/2 in. on the inside diameter to accom-
modate dimensiocnal changes of the graphite and to allow clearances for
removing and replacing the core assembly. Salt flow passages and appro-
priate orificing are provided between the graphite slabs to maintain the
temperature at acceptable levels. A system of Hastelloy N bands and verti-
cal rods key the slabs together and cause the reflector to move with the
vessel wall as the systems expand with temperature. Reflector graphite is
also provided in the top and bottom heads. The amount of fuel salt in the
radial and axial reflector regions is about 1% of the reflector volume.

The primary heat exchangers are designed for primary salt in the
tubes and secondary salt on the shell side. Each exchanger removes one-
fourth of the 2250 Mw of thermal energy and has an estimated 5900 tubes,
3/8 in. OD and about 21 1/2 ft long. One objective of the design was to
minimize the primary salt content; therefore, knurled tubing was proposed
in order to increase the heat transfer coefficient. The velocity in the
tubes is 10 ft/sec; the velocity of the secondary salt is typically 7.5
ft/sec. The overall heat transfer coefficient is estimated to be 950
Btu/hr-ft2-°F. Provisions have been made in the design for replacement
of the tube bundle.

Pumps similar to those used in the MSRE are proposed for the MSBR.

It is required that fission gases be removed from the circulating
salt for neutron econcmy reasons. This will be accomplished in a lO%
bypass flow around the heat exchanger and core. This side stream leaves
the main loop just downstream from the pump discharge. Previously in-
Jected helium bubbles containing the fission gasses will be removed by
a gas separator in the side stream. The separated gas is combined with
any overflow salt from the expansion tank and routed through a cooled
line to the primary salt drain tank. After a minimum holdup of one hour,
the gas is sent to charcoal beds for 135 Xe holdup and decay. After part
of the gas is further purified, the relatively clean helium is pressurized
and injected as bubbles into the primary salt loop bypass stream Jjust
before the side stream re-enters the main loop at the primary pump suction.

The primary system drain tank, used above for gas holdup, is sized
to hold all the primary salt plus the secondary salt which would also

drain to this tank in the event of failure of a tube in one of the primary
51

heat exchangers. The drain tank is equipped with multiple natural con-
vection salt cooling systems which transfer the afterheat to boiling water
systems in which the steam is condensed in a natural draft stack.

The heat removed from the primary heat exchangers by the secondary
salt is used to generate steam in tube and shell steam generators with
the secondary salt on the shell side. To reduce the possibility of freez-
ing of the secondary salt, the feedwater enters the steam generators at
TOO°F. Supercritical steam is generated at 1000°F and 3600 psia. There
are four 140 Mw thermsl steam generators in each of the four secondary
systems.

Part of the steam from the steam generators will be used to heat
the feedwater in a mixing tee from about 550°F to the TOO°F required.
Otherwise, the steam cycle will be similar to that of the TVA Bull Run

steam plant.
10.

11.

12.

52
REFERENCES

R. C. Briant, et al., Nuclear Science and Engineering, Vol. 2, No. 6,

pp- T795-853 (1957).

P. N. Haubenreich, et al., MSRE Design and Operations Report, Part IIT,
Nuclear Analysis, ORNL-TM-730, Oak Ridge National Laboratory, Feb. 3,
196k

 

R. B. Briggs, et al., MSR Program Semiann. Progr. Rept., July 31, 196k,
ORNL-3708, Oak Ridge National Laboratory.

P. N. Haubenreich, et al., MSR Program Semiann. Progr. Rept., Feb. 28,
1970, ORNL-45L8, pp. 125, Oak Ridge National Laboratory.

E. S. Bettis, et al., MSR Program Semiann. Progr. Rept., Feb. 28, 1969,
pp. 49-T1, USAEC Report ORNL-4396, Oak Ridge National Laboratory.

E. S. Bettis, et al., MSR Program Semiann. Progr. Rept., Aug. 31, 1969,
pp. 39-58, USAEC Report ORNL-LUL9, Oak Ridge National Laboratory.

0. L. Smith, et al., MSR Program Semiann. Progr. Rept., Aug. 31, 1968,
ORNL-L34L4, p. 71, Oak Ridge National Laboratory.

0. L. Smith, et al., MSR Program Semiann. Progr. Rept., Feb. 28, 1969,
ORNL-L396, pp. 8487, Oak Ridge National Laboratory.

O. L. Smith and J. H. Carswell, MSR Semiann. Progr. Rept., Aug. 31,
1969, ORNL-4449, pp. 69—T70, Oak Ridge National Laboratory.

R. C. Robertson, MSRE Design and Operations Report, Part I, Description
of Reactor Design, ORNL-TM-728, January 1965, Oak Ridge National
Laboratory.

 

L. E. McNeese, MSR Program Semiann. Prog. Rept., Feb. 28, 1970, USAEC
Report ORNL-454L8, pp. 277288, Oak Ridge National Laboratory.

P. N. Haubenreich and J. R. Engel, Experience with the MSRE, Nucl.
Appl. Tech., 8, 118 (1970).
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. Baes

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antor
Cardwell
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Crowley
Culler
Distefano
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FBatherly
Engel

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Fraas

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Fuller
Furlong
Gabbard

. Gallaher
. Gilpatrick
. Grimes

. Grindell
Guymon

. Harley

. Harms

. Haubenreich
. Helms
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. Hoffman

. Holz

. Huntley

. Jordan

. Kasten

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54

M. Weinberg 136. L. V. Wilson

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