NAT_MSBRfuelcycle.txt

REACTOR PHYSICS AND
FUEL-CYCLE ANALYSES

A. M. PERRY and H. F. BAUMAN
Oak Ridge National Laboratory
Oak Ridge, Tennessee 37830

Received August 4, 1969
Revised October 9, 1969

 

As presently conceived at Oak Ridge National
Laboratory and described in this issue, the single-
fluid Molten-Salt Breeder Reactor, operating on
the “2Th-P3U fuel cycle and based on a veference
design, has a breeding rvatio of ~1.06, specific
fissile inventory of 1.5 kg/MW(e), a fuel doubling
time of ~20 years, and fuel cycle costs of ~0.7
mill/kWh(e). Start-up may be accomplished with
either enrviched uranium orv plutonium, with little
effect on fuel cost; the breeding ratio, averaged
over reactor life, is reduced 0.01 to 0.02 relative
to the equilibrium cycle.

Operated as a converter, with limited chemical
processing, the veacltor may have a conversion
ratio in the vange 0.8 to 0.9 with fuel cycle costs
of 0.7 to 0.9 mill/kWh(e).

 

INTRODUCTION

One of the most important aspects of the
Molten-Salt Reactor (MSR) concept is that it is
well suited for breeding with low fuel-cycle costs,
and it does so in a thermal reactor operating on
the ***Th-?**U fuel cycle. This is true not primar-
ily because of any unique nuclear characteristics,
for the reactor is similar to other thermal reac-
tors in terms of attainable fuel-moderator ratios,
the unavoidable presence of certain parasitic neu-
tron absorbers, and reliance on a fertile blanket
to reduce neutron losses by leakage to an accept-
ably low level for breeding. Indeed, the concept
might be thought to have some a priori disadvan-
tage, because a substantial fraction of the fissile
material is invested in the heat transfer circuit
and elsewhere outside the reactor core. The pe-
culiar suitability of the molten-salt reactor for

208 NYJCLEAR APPLICATIONS & TECHNOLOGY

 

KEYWORDS: molten-salt re-
actors, fuels, economics, op-
eration, breeding, thorium-232,
uranium-233, performance
MSBR, fuel cycle, cost, breed-

ing ratio

’

economical thermal breeding stems rather from
the practical possibility of continuous removal of
fission-product wastes and ***Pa, and virtually ar-
bitrary additions of uranium or thorium, without
otherwise disturbing the fuel. This fundamental
aspect of the molten-salt reactor, details of which
are discussed in other papers of this series, has a
profound effect on the relationship between neu-
tron economy and fuel-cycle cost. The coinci-
dence of good neutron economy with low fuel-cycle
cost which characterizes the molten-salt reactor
appears to be unique among thermal reactors and
will be described more fully in this paper.

GENERAL NUCLEAR CHARACTERISTICS

The LiF/BeF; carrier salt used in the MSR
concept is not by itself a very good moderator. Its
moderating power is about half to two-thirds that
of graphite (the exact value depending on the pro-
portions of Li and Be in the salt), while its macro-
scopic absorption cross section is an order of
magnitude greater than that of graphite, even with
the feed lithium enriched to 99.995% in the “Li
isotope. (With this composition, <10% of the neu-
tron absorptions in the salt occur in °Li; nearly
half are in fluorine, and about a third in "Li.) It is
evident, therefore, that an additional moderator is
needed, and graphite is selected for this purpose
because of its compatibility with the salt.

There is only a weak connection between the
fissile fuel concentration in the carrier salt and
the heat transfer characteristics of the salt (aris-
ing primarily from the influence of the thorium
concentration on the physical properties of the
salt), and as a consequence one has considerable
latitude in selecting the uranium (and thorium)
concentrations in the salt. Because the carrier
salt itself constitutes a significant neutron poison,
the fuel concentration in the salt must not be set

VOL. 8 FEBRUARY 1970
at too low a level, but must be high enough for the
fuel to compete favorably (for neutrons) with the
lithium and the fluorine in the salt. On the other
hand, it must not be too high, lest the inventory of
fuel outside the reactor core become excessive.
The optimum fuel concentration, typically ~0.2
mole% of UF, in the salt, or ~1 kg of uranium per
cubic foot of salt, is interrelated with the neutron
spectrum in the reactor, which is a function of the
relative proportions of fuel salt and graphite mod-
erator in the core. Too large a proportion of salt
leads to an excessive fuel inventory and to a
poorly thermalized neutron spectrum, with a re-
duced neutron yield, n; too large a proportion of
graphite leads to excessive neutron-absorption
losses in the graphite. An optimum salt volume
fraction is typically found to be ~13 to 15%.

The proper balance of the above factors does,
of course, depend in part on the power density in
the reactor core, which may be selected almost
independently of the power density in the remain-
ing parts of the primary salt circuit. The maxi-
mum power density in the core is limited by fast
neutron damage to the graphite moderator, while
the removal power density in the external power
recovery circuit is limited primarily by heat
transfer and pressure-drop considerations and by
requirements for pipe flexibility in the piping runs
between the reactor vessel and the heat exchang-
ers.

The necessity for maintaining a sufficiently
high fuel concentration to suppress neutron losses
in the carrier salt and in the moderator, together
with the requirement for appreciable core size
simply to generate the requisite amount of power,
leads to the conclusion that thorium must be pres-
ent in the core, not merely in a surrounding blan-
ket. However, the question of how the thorium 1s
to be incorporated in the core is crucial to the
MSBR concept. One quickly recognizes several
distinct possibilities, some much more desirable
in principle than others, but full of implications
with respect to reactor design and chemical pro-
cessing.

We have previously given serious consideration
to a two-fluid reactor in which the fissile and
fertile materials are carried in separate salt
streams, the bred uranium being continuously
stripped from the fertile stream by the fluoride
volatility process. Blanket regions contain only
the fertile salt, while the core contains both fis-
sile and fertile streams; these streams must be
kept separate by a material with a low-neutron
cross section, that is, by the graphite moderator
itself. This approach appears to yield the best
nuclear performance, owing primarily to a combi-
nation of maximum blanket effectiveness and min-
imum fuel inventory. It also exhibits attractive

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

Perry and Bauman FUEL-CYCLE ANALYSES
safety characteristics because expansion of the
fuel salt, upon heating, removes fissile material
from the core while leaving the thorium concen-
tration unchanged. The concept does, however,
involve important questions regarding the reli-
ability of the graphite ‘‘plumbing’’ in the core, the
adequate proof of which may require a good deal
of time and testing.

The present approach employs a single salt
stream which contains both the fissile and the fer-
tile materials. This concept represents a modest
extrapolation of the technology already demon-
strated in the MSRE. A central feature of the
concept is the manner in which the single salt
composition can be made to function adequately
both in the core and in the blanket (or outer core)
regions. This is done by the simple expedient of
altering the salt volume fraction, making it con-
siderably larger in the blanket than in the core.
This undermoderation results in enhanced reso-
nance capture of neutrons by thorium in the outer
region, gives rise to a negative material buckling
in the outer region, and should in principle cause
a fairly rapid decrease in power density in the
blanket as a function of distance from the core
boundary. In practice, the distinction between the
core and blanket regions is not as clear cut as
this argument may suggest, but the idea works
reasonably well. Figure 1 illustrates the power
density distribution for our present reference de-
sign based on the single-fluid concept. The en-
hancement of resonance neutron capture in the
blanket (or outer core) region is indicated by the
ratio of neutron absorptions in ?**Th to those in
2831y, this ratio is about 1.0 in the core, and 1.3
in the blanket. The salt annulus, which is required
to allow the periodic replacement of the modera-
tor, functions as a part of the outer core region.

The principal shortcoming of the single-fluid
concept, of course, is the substantial investment
of fissile material in the blanket region. This re-
sults in a rather different compromise between
breeding gain and specific inventory than in the
two-fluid concept, leading both to reduced effec-
tiveness of the blanket region and to an apprecia-
ble increase in fuel inventory. Fortunately, this
feature of the single-fluid reactor is partly offset
by a reduction in neutron captures in the carrier
salt, owing to the fact that a single carrier salt
contains both fissile and fertile materials.

The preceding qualitative discussion is in-
tended to provide a general understanding of the
interplay of factors affecting the selection of
MSBR design parameters. These factors are of
course quite numerous. They include core size,
radial and axial blanket thickness, reflector thick-
ness, salt volume fractions in the core and blan-
ket regions, thorium and uranium concentrations,

" FEBRUARY 1970 209

 
Perry and Bauman FUEL-CYCLE ANALYSES

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70

o 60 w it = ]

o~ - L

5 N\ = &

X 50 \

© 2

= |
P(r) =

5 40 \ =

L. —

= \ =

3 AN

= \\

o 20 N

<

Ll

(e

35 6 (>50&\ \

S 10 N

) \'7

0 4

0 50 100 150 200 250 300
RADIAL DISTANCE FROM CORE CENTERLINE (cm)

Fig. 1. Radial power density and fast flux distribu-

tions—single-fluid MSBR.

chemical processing rates, and reactor power
level. Because the interaction of all these factors
is rather complex, and because of the need to
identify optimum values of the design variables
rather closely, we have found it convenient to
make use of a comprehensive, automatic reactor
optimization procedure for arriving at that combi-
nation of design parameters that will produce, in
some sense, the best attainable performance. The
Reactor Optimization and Design code (ROD) is
based on a gradient projection method for locating
the extreme value of a specified figure of merit,
which may be any desired function of the breeding
ratio, the specific fuel inventory, various ele-
ments of the fuel cycle and capital costs, or any
other factors important to the designer. The
computational procedure comprises multigroup
(synthetic), two-dimensional diffusion-theory cal-
culations of the neutron flux, an equilibrium fuel-
cycle calculation which determines the critical
fuel concentration and nuclide composition consis-
tent with processing rates and other variables,
and the gradient projection calculation for moving
the cluster of independent variables in the direc-
tion that most rapidly improves the figure of
merit. The optimization may be constrained by
limiting the allowed range of the independent vari-
ables, or by selecting in advance the desired value

210

NUCLEAR APPLICATIONS & TECHNOLOGY

(or a limiting value) of certain derived quantities,
such as the maximum power density.

The figure of merit used here in determining
reactor design specifications is related to the ca-
pability of a reactor type to conserve fuel supply
in an expanding nuclear economy. For the special
case of a linear increase in power generation, the
total amount of natural uranium that must be
mined up to the point when the system becomes
self-sufficient (i.e., independent of any external
supply of fissionable material) is proportional to
the product of the doubling time and the specific
fuel inventory. We have chosen to optimize our
MSBR design primarily on the basis of a quantity
which we call the fuel ‘“‘conservation coefficient,’’
defined as the breeding gain times the square of
the specific power, which is equivalent to the in-
verse of the product of the doubling time and the
fuel specific inventory. Therefore, a maximum
value of the conservation coefficient is sought in
the optimization procedure.

EQUILIBRIUM FUEL-CYCLE RESULTS

The result of a reactor optimization calculation
is a set of specifications for the optimum reactor
configuration, subject to any imposed constraints,
together with a complete description of its equi-
librium fuel cycle. This description includes the
multigroup neutron flux distributions, the result-
ing power distribution, and the consistent set of
concentrations of all nuclides present in the reac-
tor. We have imposed constraints on maximum
power density (i.e., minimum graphite life), on
overall reactor vessel dimensions, and on chem-
ical processing rates which we believe will result
in near-minimum power cost. Although we lack
specific information as to the cost of chemical
processing as a function of fuel processing rate
for the liquid-metal extraction process, it appears
that processing equipment sizes and operating
costs will be comparable with those for the
fluoride-volatility/uranium-distillation process
considered for the two-fluid reactor. We have
therefore fixed the processing rates, listed in
Table I, at values found to be essentially optimum
in studies of the two-fluid reactor, with minor ad-
justments appropriate to the extraction process.
While subsequent improvements in processing
cost estimates may suggest some change in opti-
mum processing rate and some change in fuel cost
estimates, we do not expect that these will result
in any major revision in performance estimates
for the reactor.

The reference reactor configuration which re-
sults from these and other (engineering) consider-
ations is described by Bettis.' A summary of its
nuclear design characteristics is given in Table L

VOL. 8 FEBRUARY 1970
Perry and Bauman FUEL-CYCLE ANALYSES

TABLE I

Characteristics of the One-Fluid MSBR Reference Design

 

B. Performance

 

A. Description
Identification CC93
Power, MW ((e) 1000
MW (th) 2250
Plant factor 0.8
Dimensions, ft
Core zone 1
Height 13.0
Diameter 14.4
Region thicknesses
Axial: Core zone 2 0.75
Plenum 0.25
Reflector 2.0
Radial: Core zone 2 1.25
Annulus 0.167
Reflector 2.5
Salt fractions
Core zone 1 0.132
Core zone 2 0.37
Plena 0.85
Annulus 1.0
Reflector 0.01
Salt composition, mole%
UF4 0.228
ThF4 12
BeFs 16
LiF 72
Processing cycle times for removal of
poisons?
1. Kr and Xe; sec 20
2. Se, Nb, Mo, Tec, Ru, Rh, Pd, Ag,
Sb, Te, Zr; sec 20
3. Pa; Cd, In, Sn; days 3
4, Y, La, Ce, Pr, Nd, Pm, Sm, Eu,
Gd; days 50
5. Sr, Rb, Cs, Ba; year 5
6. Br, I; days 5

 

Conservation coefficient, [MW (th)/ kg]2 14.3
Breeding ratio 1.062
Yield, % per annum 3.18
Inventory, fissile, kg 1478
Specific power, MW(th)/kg 1.52
Doubling time, system, year 22
Peak damage flux, E >50 keV, n/ (cm2 sec)
Core zone 1 3.2x10™
Reflector 4.2x10"
Vessel 3.7x10™
Power density, W/ cm’
Average 22.2
Peak 65.2
Ratio 2.94
Fission power fractions by zone
Core zone 1 0.765
Core zone 2 0.167
Annulus and plena 0.056
Reflector 0.012

 

 

 

 

apccording to our present flow sheet, Zr, Cd, In, and Sn
will be removed on a 200-day cycle, and Br and I on a
50-day cycle. The additional poisoning, however, is
negligible.

A neutron balance for this case is given in Table
I, in which the normalization is to one neutron
absorbed in ?**U plus **U.

Uncertainties in Neutron Cross Sections

We have estimated the effect of uncertainties in
neutron cross sections on the calculated perfor-

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

mance of the MSBR. By far the most important
effect is the uncertainty in the average value of 7
of 2**U in the MSBR spectrum, which leads to an
uncertainty in the breeding ratio of +0.012. Un-
certainties in the cross sections of other impor-
tant MSR nuclides (such as F) make a relatively
small contribution to the overall uncertainty in the
breeding ratio, which is estimated to be +0.016. A
detailed discussion of cross-section uncertainties
is given in a report by Perry.*

Equilibrium Fuel-Cycle Costs

As stated before, the molten-salt breeder re-

| actor exhibits unusually low fuel-cycle costs in

combination with good breeding performance. This
results primarily from the low specific fuel in-
ventory and from a small but non-negligible ex-
cess production of fuel, which results from the

ability to process the fuel rapidly at what appears

to be a very low unit cost.

The inventory of fissile material in the reactor
and chemical processing plant amounts to some
1480 kg, including ~100 kg each of **U and “*Pa;
when valued at $13.00/g for ?**U and **Pa and
$11.20/g for 2*°*U, this material is worth $19 mil-
lion. With an effective annual inventory charge
rate of 10%/year and a 0.8 plant factor, the fuel
inventory thus contributes 0.27 mill/kWh(e) to the

FEBRUARY 1970 , 211
Perry and Bauman

FUEL-CYCLE ANALYSES

 

 

 

TABLE II
Neutron Balance, Single-Fluid MSBR
Absorptions Fissions
2331 0.9239 0.8239
235 0.0761 0.0619
232Th 0.9853 0.0031
234 0.0817 0.0004
2:3Pa 0.0017
° 0.0088
23TNp 0.0061
°Li 0.0049
"Li 0.0159
'Be 0.0071 (0.0046)2
Bp 0.0205
Graphite 0.0519
Fission products 0.0196
Leakage 0.0276
ne 2.2311

 

 

 

 

 

a(n,2n) reaction.

fuel-cycle cost. The fuel salt, with a composition
LiF/BeF:/ThF4 = 72/16/12 mole%, respectively,
is estimated to be worth $3 million, including the
thorium. At 10%/year, this contributes 0.04 mill/
kWh(e) to the fuel cycle cost. For a conversion
ratio of 1.062, fuel production results in a reduc-
tion of 0.09 mill/kWh(e) in the fuel cycle cost.
The cost of thorium burnup, in contrast, is negli-
gible [~0.002 mill/kWh(e)].

The chemical process for removal of fission
products, which is under development for use with
the single-fluid MSBR, involves the accumulation
of rare-earth fission products in a portion of the
salt stream in the liquid bismuth extraction tower.
The concentration of rare-earth trifluorides in
this salt is limited by solubility to ~0.7 mole%; it
is presently planned to limit this concentration by
discarding ~0.5 ft°/day of carrier salt having
~100 times as high a concentration of rare earths
as the salt circulating in the reactor. The makeup
of carrier salt (including ThF,) required to com-
pensate for this discard thus contributes ~0.5 X
$1846/ft> = $932/day to the fuel cost, i.e., 0.05
mill/kWh(e) at 0.8 plant factor.

The cost of processing the fuel for removal of
fission products and for isolation of ***Pa from the
circulating salt stream is difficult to assess pre-
cisely. Our tentative estimate of these costs, in-
cluding both capital and operating expense, is ~0.3
mill/kWh(e), based on the rapid processing rates
indicated in Table I.° In summary, therefore, we
estimate that the equilibrium fuel-cycle cost will
be ~0.7 mill/kWh(e), as shown in Table III, for a
single-fluid MSBR of the reference design.

212 NUCLEAR APPLICATIONS & TECHNOLOGY

EFFECT OF CHANGES IN REACTOR
DESIGN PARAMETERS

Although our computational procedure is de-
signed to lead directly to the optimum combination
of reactor parameters, it is nonetheless a matter
of some interest to see how deviations of these
parameters from their optimum values will affect
the performance of the reactor. The influence of
these parameters on reactor performance may be
investigated by assigning specific perturbed val-
ues to each parameter in turn, the others retain-
ing their reference values, and performing the
flux and equilibrium fuel-cycle calculations for
the perturbed cases. In some instances, a se-
lected subset of the unperturbed variables may be
allowed to be reoptimized, using ROD, if there is
reason to suppose that such reoptimization will
partially compensate for any adverse effect of the
perturbation. The effect of several specific de-
partures from the reference 1000 MW(e) design
given in Table I is discussed in the following
paragraphs.

Reactor Plant Size

The effectiveness of the blanket (outer core)
region depends very much on its thickness. Nor-
mally, the blanket will contain a larger fraction of
the salt inventory for a small than for a larger
reactor. Thus, both the fuel specific power and
the breeding gain, for an optimized reactor, in-
crease as the reactor plant size is increased, as
shown in Fig. 2. This is true when the reactors
are compared at equal core life (the solid curves)
or at equal average core power density (the dashed
curves). A brief listing of dimensions and other
parameters for 500, 1000, 2000, and 4000 MW(e)
reactors is given in Table IV.

 

 

TABLE III
Equilibrium Fuel-Cycle Cost
Cost
Cost Element mills/kWhe)

Fuel inventory?2 0.27
Salt inventorya 0.04
Salt makeup 0.05
Moderator replacement 0.10
Processing 0.30
Subtotal 0.76
Fuel production credit -0.09
Total fuel-cycle cost 0.67

 

 

 

 

a Inventory charge 10% per annum.

VOL. 8 FEBRUARY 1970

 
—————— CONSTANT CORE LIFE
g | = —=——CONSTANT RATIO OF REACTOR 40
POWER TO CORE VOLUME

———=G x 100

comman_
—
-—

30
cC

5 —v, %/year | 20
L, years

3 ! 10

!

CONSERVATION COEFFICIENT

    

INVENTORY, GAIN, YIELD, CORE LIFE

I,

kg/MW(e) | O

0 ] 2 3 4 5
REACTOR POWER [103 MW(e)]

Fig. 2. Effect of power level on MSBR performance.

Graphite Moderator Life

The useful life of the graphite moderator is
limited by radiation damage effects, caused by
fast neutrons. As discussed by Eatherly,” we have
for the present adopted a limiting fast-neutron
fluence (E > 50 keV) corresponding to zero net
graphite volume change at the end of exposure.
Since the fast-neutron flux is almost entirely de-
termined by the local power density (per unit vol-

TABLE 1V

Performance of Single-Fluid MSBR’s as
a Function of Plant Size

 

 

 

 

 

 

 

 

 

 

Reactor Power, MW(e)
500 1000 2000 4000
Core height, ft 9.44 | 11.0 17.44 | 23.0
Core diameter, ft 10.42 | 14.4 19.36 | 25.5
Salt specific volume,

ft/ MW(e) 1.75 1.68 1.62 1.55
Fuel specific inventory,

kg/ MW(e) 1.65 1.47 1.36 1.28
Peak power density, W/cm? | 62.2 65.2 66.1 65.9
Peak flux (> 50 keV),

10" n/ (cm? sec) 3.04 | 3.20 | 3.25 | 3.24
Core life, years at 0.8 PF 4.3 4.1 4.0 4.0
Leakage,

n/ fissile absorption x 1000 | 3.89 2.44 1.53 0.96
Breeding ratio 1.043 1.065 1.076 1.083
Annual fuel yield, %/ year 1.99 3.34 4.28 4.95
Conservation coefficient 8.0 15.1 21.0 25.9

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

FEBRUARY 1970

Perry and Bauman FUEL-CYCLE ANALYSES
ume of salt-plus-moderator), there is a nearly
unique relationship between maximum core power
density, plant utilization factor, and useful core
life, for a specified maximum fluence. While one
expects a higher power density to be accompanied
by a reduced fuel inventory, there is in fact a
power density above which increased neutron
leakage losses and other associated losses in
breeding gain more than offset the reduced inven-
tory, and the fuel yield and the conservation coef-
ficient then decrease. These trends are exhibited
in Fig. 3, which shows breeding gain, specific in-
ventory, fuel yield, and conservation coefficient as
a function of core life. In this comparison, blan-
ket, reflector, and plenum thicknesses were held
constant, and the core size was specified. Only
the salt volume fraction was reoptimized, and it
changed very little.

Thorium Concentration

The thorium concentration in the fuel salt pri-
marily influences the uranium inventory and the
breeding ratio. For a reactor configuration very
similar to our present reference design (but
having a slightly lower estimate of the required
external salt volume) we have examined the influ-
ence of thorium concentration on reactor perfor-
mance over the range 10 to 14 mole% ThF4 Cross
sections were carefully computed for each region
in each case and iteratively adjusted to allow for
resultant changes in reactor configuration. In
these calculations, core size, radial and axial
blanket thicknesses, and core salt volume fraction
were all subject to reoptimization.

 

 

 

 

 

 

 

 

 

2 7 \\ y 12
Ne

 

10 20

| I (x 10) =

X y ’.:

8 Z 185
A gl—"\
= G (x 100) 23
<C — —
< ‘////‘ =3
2 b 16 =%
8 | A O s

2 / v
) / O >
D o
4 14 o
3 \ O
Ll — L)
— -
> <>(E

o

L)

2

S

REFERENCE DESIGN

]

 

 

 

 

 

0 10
2 3 4 5 6 7 8
GRAPHITE CORE LIFE (year)
Fig. 3. Performance of 1000 MW(e) MSBR as a func-

tion of core life (at 0.8 plant factor).

213
Perry and Bauman FUEL-CYCLE ANALYSES

The basic interplay, of course, is between ris-
ing breeding gain and rising inventory, as thorium
and uranium concentrations are increased. Both
the annual fuel yield (y) and the conservation co-
efficient (CC) should exhibit a peak, when plotted
as a function of thorium concentration, but the
peaks will occur at different places because of the
difference in weight assigned to the specific
power. These trends are shown in Fig. 4. It may
be seen that there is quite a broad maximum in
the conservation coefficient in the vicinity of 12
mole% ThF 4e

Salt Volume Fractions

Core. In all of our calculations, the optimum salt
volume fraction in the core has fallen in the range

12 to 15 vol%, with a carbon/fissile-uranium atom

ratio close to 9000. As indicated earlier, the vol-
ume fraction is rather closely determined by a
balance between fuel inventory, degree of neutron
moderation, and neutron absorptions in the mod-
erator; for the reference design, the optimum salt
fraction was 0.132.

‘Blanket. The volume fraction of salt in the blan-
ket (outer core) is central to the whole concept of
a single-fluid, 1000 MW(e) molten-salt breeder
reactor. We have tested the effect of variations in
salt fraction (in the radial blanket) under the spe-
cial assumptions of constant overall salt volume
and constant outer diameter of the blanket region.
Results of these calculations show that a broad
optimum exists in the range of 0.35 to 0.6 for the
salt fraction. The choice of 0.37 for the reference
reactor was initially selected to permit, if de-
sired, the use of a randomly packed ball bed in the
blanket.

 

 

 

 

 

 

 

 

 

 

 

 

 

8 18
o/
Z 6 g-J @
3 ___________________-—-———-—-""" G (x 100) =
o = >
g 4 14 5
- y 8%
2 __—T 1 (x10) §%
> <C .J
2 12 25
REFERENCE DESIGN 2
I °
0 ' 10
10 N 12 13 14 15

THORIUM CONCENTRATION (mole%)

Influence of thorium concentration on the per-
formance of a single-fluid MSBR.

Fig. 4.

214

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

233pa Removal Rate

Rapid and inexpensive isolation of 2**Pa from
the circulating fuel stream is essential for eco-
nomic breeder operation of the single-fluid
molten-salt reactor concept. The necessity for
removing the Pa and allowing it to decay outside
the neutron flux may be seen from Fig. 5, in which
we show the approximate loss in breeding ratio
associated with Pa absorptions as a function of
specific power and of the processing cycle time
for continuous isolation of Pa from the circulating
salt stream. (This loss includes both the neutron
losses in the ***Pa and the destruction of latent
“*U.) As one would expect, the Pa removal cycle
must be short compared to the 40-day mean decay
life, if significant losses are to be avoided. The
effect of the Pa removal rate (along with certain
fission products) on the breeding performance of
MSR’s is shown in Fig. 6. Note that for no Pa re-
moval, the reactors considered are high perfor-
mance converters rather than breeders.

Removal of Fission Products

The fission products may be divided into sev-
eral groups, according to their chemical behavior
in molten salt systems. The main groups, roughly
in order of importance as neutron poisons, are the
noble gases, the rare earths, the noble and semi-
noble metals, the volatile fluorides, and the stable
fluorides not readily separable from the carrier
salt.

 

| 1T

S = 3.0 MW(th)/k

’
//2.5
/ |

0.06 2

\%_fl

d

8
f

 

0.08

N

—

 

—

<\

 

0.04 /

// //
y

.

e

 

 

\_.

 

 

 

LOSS IN BREEDING RATIO
NN\
N
AN
' N\ N
\\
1
8——¥i

LI\ N
=

 

 

A\

-REFERENCE

R

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0.02 ,
CYCLE TIME| AT
/’/ 4 0 5 l
:,v: bt
- |
0 Ll
1 10 100 1000
PROCESSING CYCLE TIME FOR Pa REMOVAL (days)
Fig. 5. Approximate loss in breeding ratio associated

with neutron absorptions in 233Pa for a single-
fluid MSBR.

FEBRUARY 1970
 

 

 

 

 

 

 

 

 

 

1.10

1.05 } -
- 3 year SALT DISCARD ———_| 3-day Pa CYCLE
= (NOT INCLUDING Zr) ———
= 30-day Pa CYCLE
S 1.00
v
o
Ll
= 4 year
S SALT DISCARD
< 0.95 (INCLUDES Zr) 8Ye°'§\

NO Pa REMOVAL
0.90
0 100 200 300
RARE EARTH REMOVAL CYCLE (days)

Fig. 6. Conversion ratio vs processing cycle times—

single~-fluid MSBR.

An outstanding characteristic of the molten salt
reactor is the relative ease with which the noble
gases—notably '**Xe—can be removed from the
salt stream. This aspect is discussed in detail by
Scott® suffice it to say here that we estimate a
residual poisoning effect (reduction in breeding
ratio), due to '*°Xe together with other noble gas
fission products and their daughters, of not more
than 0.005.

The effect of the rare earth removal cycle on
the breeding performance of single-fluid MSR’s
(at various Pa removal rates) is shown in Fig. 6.
The ability to process the fuel rapidly for rare
earth removal is one of the most important char-
acteristics of the molten salt reactors. While
rapid rare earth and Pa removal are required for
good breeding performance, there is considerable
flexibility to trade off nuclear performance for
decreased processing rate (and processing cost)
at the discretion of the designer.

The ‘‘noble metals,”” which do not form stable
fluorides and do not remain in the fuel salt, would
have a serious adverse effect on the breeding
ratio if they simply attached themselves to the
graphite moderator in the core. The nuclides in
this group, which includes Nb, Mo, Tc, Ru, Rh,
and several others of less importance, have a
combined yield of 0.35 and, if they accumulated in
the reactor core over a long enough period of
time, would eliminate any chance of breeding.
Most of the nuclides in this group, however, have
small neutron-absorption cross sections and satu-

Perry and Bauman FUEL-CYCLE ANALYSES
whose potential poisoning effect is of real con-
cern. The chemical behavior of this group of fis-
sion products is discussed in detail by Grimes.
For the present purpose, it is sufficient to note

" that we believe ~10% of these nuclides will remain

in the core regions of the MSBR. On this basis,
the poison fraction, P(f), of all nuclides in the
eroup and the time-averaged poisoning, P(t), up
to time ¢ after start-up with clean graphite, are
shown in Fig. 7. Since we anticipate replacement
of the graphite after ~4 years, we may see from
Fig. 7 that the average reduction in breeding ratio
attributable to the accumulation of these nuclides
is 0.004. This effect is more than compensated by
the burnout of '°B in the graphite, and these ef-
fects are assumed to cancel, in our calculations.

The ‘‘semi-noble metals,”” including Zr, Cd,
In, and Sn, are assumed to be removed in an ad-
junct to the Pa removal process. The poison ef-
fect of this group is small and they could also be
removed, if necessary, by a gradual discard of
fuel salt (from which the uranium would be recov-
ered by fluorination).

The volatile fluorides of Br and I are relatively
unimportant and are removed in the fluorination

 

 

 

 

 

 

 

 

 

 

 

 

 

0.014
0.012 /
0.010 //
O
— / P(t)
<[
[a'4
€ 0.008
a /
Ll
Ll
Z  0.006 /
@ / /fi(t)
<T
I
° /
0.004 /
0.002 /
0

 

0 2 4 6 8 10
TIME (calendar years)

rate very slowly, or else they have very low  Fig. 7. Change in breeding ratio due to noble-metal
yields. It is primarily Mo, *Tc, '*Ru, and 1%Rh fission products in MSBR,
NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 215
Perry and Bauman FUEL-CYCLE ANALYSES
step of the Pa removal process. The stable fluo-
rides of Rb, Sr, Cs, and Ba are removed by the
discard of salt from the rare earth process on a
long cycle, typically eight years.

REACTOR START-UP AND APPROACH
TO EQUILIBRIUM

The preceding discussion has dealt only with
the equilibrium fuel cycle, in which all nuclide
concentrations—and particularly those of the ura-
nium isotopes—are time-independent and have
their asymptotic values. This condition is in fact
approached rather rapidly in an MSBR, because of
its relatively high specific power. Nonetheless, it
is of interest to examine the transient fuel-cycle
characteristics of the reactor when it is initially
fueled with enriched uranium (*®U) or with plu-
tonium.

There is considerable latitude in specifying the
initial fuel loading for the reactor, even when the
equilibrium fuel cycle is well defined, since the
initial thorium concentration may, at the discre-
tion of the operator, be quite different from the
equilibrium value. We have, in fact, tried dif-
ferent initial thorium concentrations, and find no
advantage in choosing a value different from the
equilibrium one. In the following discussion,
therefore, it will be understood that the thorium
concentration is held constant at 12 mole%.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2.0
1.6
o
L
:CTD’ 233U
> 1.2 >
(o o
O
|
=
L
=
=
= 0.8
=
J
—
w
b
) 23‘+U
0.4 T
v<k 236
N —————————
0 235U
0 4 8 12 16 20
EXPOSURE TIME — EQUIVALENT FULL-POWER YEARS
Fig. 8. Uranium isotope inventories—MSBR start-up

with enriched uranium.

216

NUCLEAR APPLICATIONS & TECHNOLOGY

25U Start-Up

Start-up of the reactor on enriched uranium
(93% **U) results in an initial conversion ratio
substantially less than unity. However, because
3% is much more reactive than 2*°U in the MSBR
spectrum, the export of fissile material from the
reactor plant may begin quite soon after start-up,
and indeed well before the conversion ratio actu-
ally exceeds unity.

Time-dependent inventories of the uranium
isotopes are shown, for the reference single-fluid
1000 MW(e) MSBR, in Fig. 8, while the conversion
ratio and the cumulative net feed of fissile mate-
rial to the circulating fuel salt are shown in Fig.
9. It may be noted that very little additional fuel
is supplied to the reactor in excess of the initial
critical mass required for start-up. There is, of
course, some loss in performance, averaged over
the life of the reactor, associated with this initial
phase of the fuel cycle. The average reduction in
breeding ratio, over the life of the plant, is
~0.018, while the increase in the present value of
the 30-year fuel-cycle cost is +0.02 mill/kWh(e).

Pu Start-Up

A useful attribute of the molten-salt reactor is
the capacity of the LiF/BeF,;/ThF, carrier salt to
contain plutonium in the form of PuFj with a sol-
ubility in excess of 1 mole% at the operating tem-
peratures of the MSBR. We have considered the
start-up of the MSBR on plutonium whose compo-
sition is typical of the material discharged from a

 

 

 

 

 

 

 

 

 

 

 

 

 

2.0 1.2
2 1.6 \\ - 1.1
o N
a /’X\ -
w2 102
L)
— =
3 ~ :
— o
“ 0.8 / o 0.9 W
L \ =
: O
<C
=
§ 0.4 0.8

%0 4 8 12 16 20 07

EXPOSURE TIME — EQUIVALENT FULL-POWER YEARS

Conversion ratio and cumulative fissile feed—
enriched uranium start-up.

Fig. 9.

VOL. 8 FEBRUARY 1970
PWR or BWR with a fuel exposure in the neigh-
borhood of 20 000 MWd/T, specifically, 59.7%
29py, 24.1% 2%°Pu, 12.2% **'Pu, and 4.0% “*Pu. A
notable difference between the plutonium and ura-
nium start-up cases is the much lower initial in-
ventory of plutonium, which is due to the very
large effective cross sections of the plutonium
isotopes in the MSBR spectrum. These isotopes
burn out very rapidly, however, and additional
plutonium must be supplied for a time until nearly
enough 23°U is produced to sustain the chain re-
action by itself. Thereafter, the concentrations of
239py  29py, and **'Pu decrease rather sharply,
while that of 22Pu decreases more slowly until the
plutonium ceases to make a net positive contribu-
tion to the reactivity. At this point, the plutonium
(now mainly ®**Pu) is isolated from the circulating
fuel salt and discarded. If this is done, the result
is a self-sustaining 2**U-fueled reactor in which
the 2%°U and 2®*U concentrations are temporarily
below their equilibrium values, and the asymptotic
breeding ratio is approached from the high side.

 

 

 

 

 

 

 

 

 

 

 

 

 

6

5
= 4
¥4
Y
W
L
S 3
'—
=
ud
=
=
Ll
>
> 2 y

241
40
| / \\\\\
239 \\
0
0 1 2 3 4

OPERATING TIME — EQUIVALENT FULL-POWER YEARS

Fig. 10. Plutonium isotope inventories—MSBR start-up
- with PWR plutonium.

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

FEBRUARY 1970

Perry and Bauman FUEL-CYCLE ANALYSES

These features of the plutonium start-up are
exhibited in Figs. 10, 11, and 12.

The average breeding ratio, over a 30-year
plant life at 0.8 plant factor, is reduced by ~0.008
relative to the equilibrium cycle. The present
value of the fuel-cycle cost is actually decreased
by ~0.04 mill/kWh(e) if one assigns to the fissile
plutonium a value 7 that of enriched *U.

OPERATION WITH LIMITED FUEL PROCESSING

The combination of good breeding performance
with low fuel-cycle cost, which is associated with
economical processes for the rapid removal of
fission products and protactinium, is an impor-
tant feature of the molten-salt reactor concept.
Nevertheless, quite satisfactory fuel costs can be
achieved in the absence of fuel processing for re-
moval of protactinium or fission products, al-
though the reactor will of course not breed when
operated in this fashion.

In examining this alternate mode of operation,
we postulate the occasional batch discard and re-
placement of the entire inventory of carrier salt,
including the thorium, but not including the ura-
nium; the latter is to be recovered by fluoride
volatility (as was done with the MSRE fuel salt)
and recycled to the reactor. The removal of
noble-gas fission products from the salt by gas
sparging and of noble-metal fission products by

2.0

 

1.6

 

1.2

 

233

 

0.8

SYSTEM INVENTORIES (10° kg)

Pu 234

— 235

0 4 8 12 16 20
OPERATING TIME — EQUIVALENT FULL-POWER YEARS

 

0.4

 

 

 

 

 

 

 

 

 

 

 

Fig. 11. Inventories of uranium isotopes and total plu-
tonium for plutonium start-up of MSBR.

217
Perry and Bauman FUEL-CYCLE ANALYSES

2.0 1.3

 

—
o

 

1.2

 

Pu PURCHASED (239 + 241)

pd

0.9
/URANIUM SOLD (233 + 235)

0 0.8
0 4 8 12 16 20

OPERATING TIME — EQUIVALENT FULL-POWER YEARS

—
.
N

—

—

 

 

 

CONVERSION RATIO

I

o
o]

 

1.0

CUMULATIVE AMOUNTS (103 kg)
CONVERSION RATIO

 

o
Y
‘

 

 

 

 

 

 

 

Fig. 12, Cumulative purchases and sales of fissile ma-
terial and conversion ratio MSBR with plutoni-
um start-up,

deposition and by escape to the off-gas system, as
observed in the MSRE, are assumed to occur.

An optimum salt discard rate exists, for which
the cost of replacing the salt (and recovering ura-
nium) is balanced against the fuel makeup cost,
which depends on the conversion ratio and hence
on the level of fission-product poisoning. In addi-
tion, however, there is a minimum discard rate
required to limit the concentration of rare-earth
trifluorides in the circulating fuel salt to an ac-
ceptable level (<1 mole%). This consideration per
se will place a limit of ~10 years on the salt re-
placement interval.

We have investigated the trend in fuel-cycle
costs, as a function of the salt replacement inter-
val, for a molten-salt converter reactor which is
defined in terms of its general nuclear character-
istics, such as neutron leakage and fuel inventory,
but whose engineering design has not been speci-
fied in detail. We found the fuel costs to be rather
insensitive to the salt replacement interval from
3 years to 8 or 10 years, with a broad minimum at
O to 6 years. The conversion ratios ranged from
0.84 for a 4-year salt-replacement interval to
0.78 for a 10-year interval.

We have not yet made any detailed estimates of
costs for the fluorination plant and for the chem-
ical treatment plant necessary to control the com-
position of the salt over long periods of time.
Allowing 0.1 mill/kWh(e) for this equipment and
its operation, and 0.1 mill/kWh(e) for graphite
replacement in the reactor core, we estimate the

218 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

fuel-cycle costs for the converter reactor, with a
salt discard cycle, to be 0.7 to 0.8 mill/kWh(e).
Compared with the MSBR, the principal cost dif-
ferences are a saving of ~0.2 mill/kWh(e) in fuel
processing, an increase of 0.2 to 0.3 mill/kWh(e)
for fuel burnup, and possibly some reduction of
fuel and salt inventory charges [<0.1 mill/kWh(e)].
Thus, the fuel-cycle cost for the converter re-
actor, without chemical processing for removal of
protactinium or of fission products, appears to be
within 0.1 mill/kWh(e) of the cost for the breeder
reactor.

TEMPERATURE COEFFICIENTS OF REACTIVITY

Expansion of the single fissile-fertile salt in
the one-fluid reactor reduces the density of most
of the absorbing materials in the same proportion,
so that one might expect a very small prompt
temperature coefficient of reactivity. The density
coefficient of the salt is in fact very small. In
addition to this, however, there is both a positive
contribution to the salt-temperature coefficient
associated with the shift in thermal-neutron spec-
trum with increasing salt temperature, and a neg-
ative contribution associated with the Doppler
broadening of resonance capture lines in thorium.
The latter predominates, resulting in a prompt
negative coefficient of -2.4 X 107 6k/k/°C.

The graphite moderator contributes a positive
component to the overall temperature coefficient,
attributable to an increase in the relative cross
section of ?*U with increasing neutron tempera-
ture. This results in an overall coefficient which
is very small, though apparently negative, i.e.,
-0.5 X 107°/°C.

The prompt negative salt coefficient will large-
ly govern the response of the reactor for tran-
sients whose periods are several seconds or less.
The small overall coefficient will provide little
inherent system response to impressed reactivity
changes, and it will consequently be necessary to
provide control rods to compensate any reactivity
changes of intermediate duration. Long-term re-
activity effects, such as those associated with the
fuel cycle, are compensated by adjustment of the
fuel concentration in the salt.

SUMMARY AND CONCLUSION

While a full economic optimization of the
single-fluid molten-salt breeder reactor has not
yet been undertaken, it is apparent that good
breeding performance can be achieved in conjunc-
tion with unusually low fuel-cycle costs, subject to
the successful completion of chemical processing

| developments now under way. That is, a breeding

FEBRUARY 1970
ratio of >1.06, together with a specific fuel inven-
tory of <1.5 kg/MW(e), will be attainable with
fuel-cycle costs of ~0.7 mill/kWh(e). The asso-
ciated fuel doubling time is ~20 years.

The MSBR fuel cycle may be initiated either
with enriched uranium or with plutonium dis-
charged from light-water reactors. The reduction
in breeding ratio, averaged over the life of the re-
actor, is <0.02 and 0.01, respectively, for the
uranium and plutonium start-up cases relative to
the equilibrium case. The present value of the
fuel-cycle cost is increased, relative to that for
the equilibrium cycle, by ~0.02 mill/kWh(e) for
the uranium-start-up case,while for the plutonium-
start-up case the cost appears to be 0.04 mill/kWh
less, based on a fissile plutonium value equal to
% that of **°U.

Development of the technology for molten-salt
reactors themselves and for the associated chem-
ical processes for removal of protactinium and
fission products need not necessarily be carried
out on precisely the same time schedule, since
an economically attractive fuel cycle may be
achieved with a salt-discard cycle, resulting in a
conversion ratio of 0.8 to 0.9. A given reactor,
operated initially in this way, could subsequently
be operated as a breeder by the addition of ap-
propriate processing equipment to the reactor
plant.

It is thus apparent that there is considerable
latitude in the mode of operation of molten-salt
reactors, with only small variations in fuel-cycle
cost. We believe that this can facilitate the devel-

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

FEBRUARY 1970

Perry and Bauman FUEL-CYCLE ANALYSES
opment of the molten-salt reactor and of the asso-
ciated chemical-processing technology on time
schedules appropriate to each, and will encourage
an orderly progress toward the achievement of an
economical breeder reactor.

ACKNOWLEDGMENTS

The authors wish to acknowledge the contributions
made by many of their colleagues to the work reported
here, and in particular those made by R. S. Carlsmith,
W. R. Cobb, E. H. Gift, and O. L. Smith, This research
was sponsored by the U.S. Atomic Energy Commission
under contract with the Union Carbide Corporation.

REFERENCES

1. E. S. BETTIS and R. C. ROBERTSON, ‘‘The Design
and Performance of a Single-Fluid MSBR,”’ Nucl. Appl.
Tech., 8, 190 (1970).

2. A. M. PERRY, ‘“Influence of Neutron Data in the
Design of Other Types of Power Reactors,”” ORNL-TM-
2157, Oak Ridge National Laboratory (March 8, 1968).

3. M. E. WHATLEY, L. E. McNEESE, W. L. CARTER,
L. M. FERRIS, and E. L. NICHOLSON, ‘‘Engineering
Development of the MSBR Fuel Recycle,”” Nucl. Appl.
Tech., 8, 170 (1970),

4, E. P. EATHERLY and D. SCOTT, ‘Graphite and
Xenon Behavior and Influence on MSBR Design,”’ Nucl.
Appl. Tech., 8, 179 (1970).

5. W. R. GRIMES, ‘“Molten-Salt Reactor Chemistry,”’
Nucl. Appl. Tech., 8, 137 (1970).

219