NAT_MSRchemistry.txt

MOLTEN-SALT REACTOR CHEMISTRY

W. R. GRIMES Oak Ridge National Labovatory,
Oak Ridge, Tennessee 37830

Received August 4, 1969
Revised October 7, 1969

 

This document summarizes the lavge program
of chewmical research and development which led
to selection of fuel and coolant compositions for
the Molten-Salt Reactor Experiment (MSRE) and

for subsequent reactovs of this type. Chemical

‘behavior of the LiF-BeF,-ZvF,~-UF, fuel mixture
and behavior of fission products duving powev op-
evation of MSRE are presented. A discussion of
the chewmical veactions which show promise for
recovery of bred ***Pa and for vemowval of fission
product poisons from a molten-salt breeder rveac-
tor is included.

 

INTRODUCTION

A single-fluid molten-salt thermal breeder
(MSBR) of the type described by Rosenthal et al.,’
Bettis and Robertson,” and Perry and Bauman®
makes very stringent demands upon its fluid
fuel.*”® This fuel must consist of elements having
low capture cross sections for neutrons typical of
the energy spectrum of the chosen design. The
fuel must dissolve more than the critical concen-
tration of fissionable material (**°U, *°°U, or
23%9py), and high concentrations of fertile material
(?32Th) at temperatures safely below the tempera-
ture at which the salt leaves the MSBR heat
exchanger. The mixture must be thermally stable,
and its vapor pressure needs to be low over an
operating temperature range (1100 to 1400°F)
sufficiently high to permit generation of high
quality steam for power production. The fuel
mixture must possess heat transfer and hydro-
dynamic properties adequate for its service as a
heat-exchange fluid. It must be relatively non-

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

FEBRUARY 1970

 

KEYWORDS: fused salt fuel,

chemical reactions, reactors,
beryllium fluorides, zirconium
fluorides, lithium fluorides,
vranium hexafluoride, repro-
cessing, protactinium, separa-
tion processes, breeding,

fission products, MSRE

aggressive toward some otherwise suitable ma-
terial of construction and toward some suitable
moderator material. The fuel must be stable
toward reactor radiation, must be able to survive
fission of the uranium (or other fissionable ma-
terial) and must tolerate fission product accumu-
lation without serious deterioration of its useful
properties. We must also be assured of a gen-
uinely low fuel-cycle cost; this presupposes a
low-cost fuel associated with inexpensive turn-
around of the unburned fissile material, and
effective and economical schemes for recovery of
bred fissile material and for removal of fission-
product poisons from the fuel.

A suitable secondary coolant must be provided
to link the fuel circuit with the steam-generating
equipment. The demands imposed upon this cool-
ant fluid differ in obvious ways from those im-
posed upon the fuel system. Radiation intensities
will be markedly less in the coolant system, and
the consequences of uranium fission will be ab-
sent. The coolant salt must, however, be com-
patible with metals of construction which will
handle the fuel and the steam; it must not undergo
violent reactions with fuel or steam should leaks
develop in either circuit. The coolant should be
inexpensive, possessed of good heat-transfer prop-
erties, and it should melt at temperatures suit-
able for steam cycle start-up. An ideal coolant
would consist of compounds which would be easy
to separate from the valuable fuel mixture should
they mix as a consequence of a leak.

This report presents, in brief, the basis for
choice of fuel and coolant systems which seem
optimum in light of these numerous—and to some
extent conflicting—requirements.

CHOICE OF FUEL COMPOSITION

Compounds which are permissible major con-
stituents of fuels for single-fluid thermal breeders

137
Grimes MSBR CHEMISTRY

are those which can be prepared from beryllium,
bismuth, boron-11, carbon, deuterium, fluorine,
lithium-7, nitrogen-15, oxygen, and the fissionable
and fertile materials. As minor constituents one
might tolerate compounds containing the other
elements in Table L

Many chemical compounds can be prepared
from the several ‘‘major constituents’’ listed
above, Most of these, however, can be eliminated
after elementary consideration of the fuel re-
quirements.””® No hydrogen- (or deuterium-)
bearing compounds possess overall properties
that are practical in such melts. Carbon, nitro-
gen, and oxygen form high melting binary com-
pounds with the fissionable and fertile metals;
these compounds are quite unsuitable as constit-
uents of liquid systems. The oxygenated anions
either lack the required thermal stability (i.e.,
NO;™ or NO; ) or fail as solvents for high concen-
trations of thorium compounds (i.e., CO;=). It
quickly develops, therefore, that fluorides are the
only suitable salts indicated in this list of ele-
ments.

Fluoride ion is capable of appreciable neutron
moderation, but this moderation is by itself in-
sufficient for good neutron thermalization. An
additional moderator is, accordingly, required.

TABLE 1

Elements or Isotopes Which may be Tolerable
in High-Temperature Reactor Fuels

 

 

Absorption Cross Section

Material (barns at 2200 m/sec)
Nitrogen-15 0.000024
Oxygen 0.0002
Deuterium 0.00057
Carbon 0.0033
Fluorine 0.009
Beryllium 0.010
Bismuth 0.032
Lithium-7 0.033
Boron-11 0.05
Magnesium 0.063
Silicon 0.13
Lead 0.17
Zirconium 0.18
Phosphorus 0.21
Aluminum 0.23
Hydrogen 0.33
Calcium 0.43
Sulfur 0.49
Sodium 0.53
Chlorine-37 0.56
Tin 0.6
Cerium 0.7
Rubidium 0.7

 

 

 

 

138

NUCLEAR APPLICATIONS & TECHNOLOGY

The only good moderator material truly compati-
ble with molten-fluoride fuel mixtures is graph-
ite,*”°

Phase Behavior Among Fluorides

Uranium tetrafluoride and uranium trifluoride
are the only fluorides (or oxyfluorides) of uranium
which appear useful as constituents of molten-
fluoride fuels. Uranium tetrafluoride (UF,) is
relatively stable, nonvolatile, and largely non-
hygroscopic. It melts at 1035°C (1895°F), but this
freezing point is markedly depressed by useful
diluent fluorides. Uranium trifluoride dispropor-
tionates at temperatures above ~ 1000°C by the
reaction

4UFs;= 3UF, + U° . (1)

It is unstable®’ at lower temperatures in most
molten-fluoride solutions and is tolerable in reac-
tor fuels only with a large excess of UF4 so that
the activity of U° is so low as to avoid appreciable
reaction with moderator graphite or container
metal. |

Thorium tetrafluoride (ThF4) is the only known
fluoride of thorium. It melts at 1111°C (2032°F)
but fortunately its freezing point is markedly de-
pressed by fluoride diluents which are also useful
with UF4.

Consideration of nuclear properties alone leads
one to prefer as diluents the fluorides of Be, Bi,
'Li, Mg, Pb, and Zr in that order. Equally simple
consideration of the stability of these fluorides®®
toward reduction by structural metals, however,
eliminates the bismuth fluorides from considera-
tion. This leaves BeF:and 'LiF as the preferred
diluent fluorides. Phase behavior of systems
based upon LiF and BeF:as the major constitu-
ents, has, accordingly, been examined in detail.™
Fortunately for the molten fluoride reactor con-
cept, the phase diagrams of LiF-BeF.-UF,; and
LiF-BeF:-ThF4 are such as to make these ma-
terials useful as fuels.

The binary system LiF-BeF: has melting points
below 500°C over the concentration range from 33
to 80 mole% BeF..'”'" The phase diagram, pre-
sented in Fig. 1, is characterized by a single
eutectic (52 mole% BeF:, melting at 360°C) be-
tween BeF, and 2LiF-BeF:. The compound
2LiF-BeF, melts incongruently to LiF and liquid
at 458°C. LiF.BeF: is formed by the reaction of
solid BeF. and solid 2LiF-BeF: below 280°C.

The phase behavior of the BeF:-UF.'”' and
BeF,- ThF,'” systems are very similar. Both sys-
tems show simple single eutectics containing very
small concentrations of the heavy metal fluoride.
ThF4 and UF4 are isostructural; they form a con-
tinuous series of solid solutions with neither
maximum nor minimum.

VOL. 8 FEBRUARY 1970
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Grimes MSBR CHEMISTRY
900 ’
848
800
~ 700 A\,
LiF+ LIQUID
o
> 600
555
% \ N
—
E:IJ \ .__/
< s00 P
w
~ 458 /
. BeF, (HIGH QUARTZ TYPE)
\ / +L1QUID
400
\\/ 360
LiF+Li2BeF4
Li,Bef,+BeF, (HIGH QUARTZ TYPE)
1 ' | |
300 < ! | 1 280
m .
@ Li,BeF, » | LiBeF; + BeF, (HIGH QUARTZ TYPE)
5 + ® LiBeFy+ BeF, (LOW QUARTZ TYPE)\ =~ | 220
LiBeF. @ ; 1
I 3 -4 | | I N |
200
LiF 10 20 30 40 50 60 70 80 90 BeF,
BeF, (mole %)
Fig. 1. The system LiF-BelF,.

The binary diagrams LiF-UF" and LiF-ThF,"
are generally similar and much more complex
than the binary diagrams discussed immediately
above. The LiF-UF, system shows three com-
pounds (none are congruently melting) and a single
eutectic, at 27 mole% UF4, melting at 490°C. The
LiF-ThF4 system contains four binary compounds,
one of which (3LiF-ThF4) melts congruently, with
two eutectics at 570°C and 22 mole% ThF, and at
560°C and 29 mole% ThFy.

The ternary system LiF-ThF4-UF,," shown in
Fig. 2, shows no ternary compounds and a single
eutectic freezing at 488°C with 1.5 mole% ThF,
and 26.5 mole% UF,;. Most of the area on the dia-
gram is occupied by the primary phase fields of
the solid solutions UF4-ThF,, LiF-4UF4-LiF-4ThF,,
and LiF-UF,-LiF-ThF,. Liquidus temperatures
decrease generally to the LiF-UF4 edge of the
diagram.

The single-fluid molten-salt breeder fuel will
need a concentration of ThFs much higher than
that of UF4. Accordingly, the phase behavior of
the fuel will be dictated by that of the LiF-BeF2-
ThF, system. Figure 3 gives the ternary system
LiF-BeF.-ThF,, this system shows a single ter-
nary eutectic at 47 mole% LiF and 1.5 mole% ThF,,
melting at 360°C.'>'" The system is complicated
to some extent by the fact that the compound

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

FEBRUARY 1970

3LiF'ThF, can incorporate Be”’?' ions in both
interstitial and substitutional sites to form solid
solutions whose compositional extremes are rep-
resented by the shaded triangular region near
that compound. Liquidus temperatures < 550°C
(1022°F) are available at ThF, concentrations as
high as 22 mole%. The maximum ThF4 concen-
tration available at liquidus temperatures of 500°C
(932°F) is seen to be just above 14 mole%. Inspec-
tion of the diagram reveals that a considerable
range of compositions with > 10 mole% ThF,4 will
be completely molten at or below 500°C.

As expected from the general similarity of
ThF, and UFs—and especially from the substitu-
tional behavior shown by the LiF-UF4-ThF4 sys-
tem (Fig. 2)—substitution of a small quantity of
UF,s for ThF. scarcely changes the phase be-
havior. Accordingly, and to a very good approxi-
mation, Fig. 3 represents the behavior of the
LiF-BeF,-ThF4-UF4 system over concentration
regions such that the mole fraction of ThF, is
much greater than that of UF4.

Oxide Fluoride Equilibria

Phase behavior of the pure fluoride system
LiF-BeF,-ThF;-UF,, as indicated above, is such
that adequate fuel mixtures seem assured. The

139
Grimes MSBR CHEMISTRY

ThF4
114

TEMPERATURE IN °C
COMPOSITION IN mole %

LiF- 4ThF,

LiF- 2ThF4

P 8I7 AN

LiF-ThE A

P 762

P 609

P 597
£ 568

3Lil—'-ThF4

(b)
&
>
©
D
E 565

{\ 5

0

LiF

848 P 610

4LiF-UF;” P500" £ 490

(.

Fig. 2.

behavior of systems such as this, however, is
markedly affected by appreciable cbncentrations
of oxide ion.

When a melt containing only LiF, BeF., and
UF4 is treated with a reactive oxide (such as H, 0O)
precipitation of transparent ruby crystals of UO2, 00
results.”” If the melt contains, in addition, an
appreciable concentration of ZrF, the situation is
markedly altered. ZrO; is less soluble than is
UO: in such melts, and the monoclinic ZrO: (the
form stable below ~ 1125°C) includes very little
UO: in solid solution. Thus, inadvertent oxide
contamination of a LiF-BeF;-ZrFs-UF, melt
yields monoclinic ZrO. containing 250 ppm of
UO:." Precipitation of cubic UO. (containing a
small concentration of ZrO.) begins only after
precipitation of ZrO, had dropped the ZrF, con-
centration to near that of the UF,.

Slow precipitation of UO. followed by a sudden
entrance of this material into the reactor core
could result in undesired increased in reactivity.
This possibility was assumed to represent a dan-
ger to the Molten-Salt Reactor Experiment. Ac-
cordingly, the MSRE fuel was chosen to contain 5
mole% of ZrF, to eliminate such a possibility.

140

(d)\@ =
o, \% Q
S Q
= % o & 6:5\ £ 500
b 5\ 52 W\
3 \ \ \\ £488
1 \l ¥
F

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

PRIMARY-PHASE AREAS

a) UF,~ThE, (ss)

b) LiF-4UF,~LiF-4ThF, (ss)

) LiF-2Th(U)F, (ss)
) TLiF-BUF,~7LiF-6ThF, (ss)
) 3LiF-Th(U)F, (ss)

(
(
(c
(d
(e
(f) LiF

©
o

\V4

PT75

Uy

LiF-4UF4 1035

iF-UF,

The system LiF-ThF,-UF,

When a mixture of LiF and BeF, containing
ThF, and UF4 is treated with a reactive oxide a
homogeneous cubic phase is produced; this phase
is a solid solution of UO; and ThQ. which is very
rich in UO.." Careful studies have shown that the
reaction

ThO:(ss) + Uy = UOa(ss) + Thip (2)
[where the subscripts (ss) and (f) indicate the
solid-phase solid solution and the molten-fluoride
solution, respectively| approaches equilibrium with
reasonable speed. Values for the equilibrium
quotient @ for this reaction

4 +

NUO 2(ss_). NTh (f) (3)

+
NTRO o(ss)’ N4U(f)

 

Q

increase with UO; concentration of the oxide phase
and decrease markedly with temperature. Since
values of @ for mixtures similar to those chosen
as fuel compositions are typically 300 to 1000, it
is clear that oxide contamination of such salts will
selectively precipitate the uranium.

FEBRUARY 1970
LiF-ThFa
3LiF-ThF4 SS

Grimes MSBR CHEMISTRY

The, 4114

   

TEMPERATURE IN °C
COMPOSITION IN mole %

 

BeF.
848 2LiF- BeFZ/scBo?zxso 400! 400 450 500 548
P 458 £ 360
Fig. 3. The system LiF-BeF;-ThFs.

It is likely, though not certain, that addition of
some ZrF, would afford protection of the sort ob-
tained with the MSRE fuel. Such addition is un-
desirable, however, since the presence of ZrF,
would certainly complicate the separation pro-
cesses described later in this paper and else-
where in this series.

The successful operation of the MSRE over a
three year period (discussed later) lends con-
fidence that oxide contamination of the fuel system
can be kept to adequately low levels. This confi-
dence, when added to the prospect that the breeder
fuel will be reprocessed (and its oxide level re-
duced) at regular intervals, suggests very strongly
that successful operation can be achieved without
added ‘‘oxide protection.’’

Tolerance levels for oxide concentration in
LiF-BeF.-UF4 and LiF-BeF:-ZrF,-UF, systems
have been studied in detail and are relatively well
understood.’®**~® Analogous values for the LiF-
BeF,-ThF4-UF, system are still largely lacking.
It is known, however, that processing of these
quaternary melts with anhydrous HF and H: serves
to remove oxide to a level below that required for
precipitation of the solid solutions. There seems
little doubt, therefore, that initial processing of

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

the type used for MSRE fuels can be successfully
applied on a large scale to the LiF-BeFz:-ThF4-
UF4 system.

MSRE and MSBR Fuel Compositions

The fuel chosen for operation of MSRE (with a
23517_238y jsotopic mixture containing 33% of the
fissionable isotope) was a mixture of 'LiF, BeFz,
ZrFs, and UF, consisting of 65, 29.1, 5, and 0.9
mole%, respectively. The uranium concentration
was fixed at ~1% so that there was less possibility
of fissile U precipitating (~ 0.3 mole% **°U was
necessary to achieve criticality and to provide a
small excess of fissionable material for power
operation of the machine). The ZrF4 was added,
as indicated above, to preclude possible inad-
vertent precipitation of UOz. Beryllium fluoride
is an extremely viscous material; its viscosity is
markedly lowered by addition of LiF. The ratio of
LiF to BeF: in the MSRE fuel was chosen to opti-
mize the conflicting demands for low viscosity and
a low liquidus temperature for the molten fuel.

The single-fluid breeder requires a high con-
centration of ThF4; concentrations near 12
mole% seem to be reasonable for good reactor

FEBRUARY 1970 141
Grimes MSBR CHEMISTRY

performance. Criticality estimates suggest that
such a fuel could be made critical in a practicable
reactor with somewhat < 0.3 mole% 2**UF,. The
ratio of "LiF to BeF, should, to decrease vis-
cosity, be kept at a value as high as is practicable.
If the liquidus temperature is to be kept at or
below 500°C (932°F) for a melt with 12 mole% of
ThF4, the beryllium concentration limits range
from 16 to 25 mole%. The most likely choice for
the MSBR fuel—and the present design composi-
tion—is, accordingly, "LiF-BeF,-ThF.-UF, at
71.7-16-12-0.3 mole%, respectively.

Choice of Coolant

The secondary coolant is required to remove
heat from the fuel in the primary heat exchanger
and to transport this heat to the power generating
system. In the MSBR the coolant must transport
heat to supercritical steam at minimum tempera-
tures only modestly above 7T00°F. In the MSRE the
heat was rejected to an air cooled radiator at
markedly higher salt temperatures.

The coolant mixture chosen for the MSRE and
shown to be satisfactory in that application is
BeF: with 66 mole% of "LiF. Use of this mixture
would pose some difficulties in design of equip-
ment for the MSBR since its liquidus temperature
is 851°F; moreover, it is an expensive material.
The eutectic mixture of LiF with BeF, (48
mole% LiF) melts at near 700°F (see Fig. 1) but it
1s relatively viscous and is expensive, especially
if "LiF is used.

The alkali metals, excellent coolants with real
promise in other systems, are undesirable here
since they react vigorously with both fuel and
steam. Less noble metal coolants such as Pb° or
Bi’ undergo no violent reactions, but they are not
compatible with Hastelloy-N, the Ni-based alloy
used in MSRE and intended for use in MSBR'’s.

Several binary chloride systems are known to
have eutectics melting below (in some cases much
below) 700°F.” These binary systems do not,
however, appear especially attractive since they
contain high concentrations of chlorides (TICI,
ZnClz, BiCls, CdClz, or SnClz), which are easily
reduced and, accordingly, corrosive; or chlorides
(AlCls, ZrCl, HICls, or BeCly) which are very
volatile. The only low-melting binary systems of
stable, non-volatile chlorides are those containing
LiCl; LiCl-CsCl (330°C at 45 mole% CsCl), LiCl-
KC1 (355°C at 42 mole% KCl), LiCl-RbCl (312°C at
45 mole% RbCl). Such systems would be relatively
expensive if made from "LiCl, and they could lead
to serious contamination of the fuel if normal LiCl
were used.

Very few fluorides or mixtures of fluorides are
known to melt at temperatures below 370°C. Stan-

142 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

nous fluoride (SnF:) melts at 212°C. This ma-
terial is probably not stable during long term
service in Hastelloy-N; moreover, its phase dia-
grams with stable fluorides (such as NaF or LiF)
show high melting points at relatively low alkali
fluoride concentrations. Coolant compositions
whichwill meet the low liquidus temperature spec-
ification may be chosen from the NaF-BeF, or
NaF-LiF-BeF, system. These materials are al-
most certainly compatible with Hastelloy-N, and
they possess adequate specific heats and low
vapor pressures (discussed in the next section).
They (especially those including LiF) are moder-
ately expensive, and their viscosities at low
temperature are certainly higher than desirable.
It is possible that substitution of ZrF,; or even
AlF; for some of the BeF. would provide liquids
of lower viscosity at no real expense in liquidus
temperature.

It now appears that the best choice for the
MSBR secondary coolant is the eutectic mixture of
sodium fluoride and sodium fluoroborate. The
binary system NaF-NaBF, has been described as
showing a eutectic at 60 mole% NaBF, melting at
304°C (580°F).*“?? Studies here have shown that
this publication is seriously in error. Boric oxide
substantially lowers the freezing point of NaF-
NaBF; mixtures; the original authors may well
have used quite impure materials. Use of pure
NaF and NaBF, leads to the conclusion that the
system shows a single eutectic at 8 mole% NaF
and a melting point of 385°C (725°F),2® as shown in
Fig. 4.

At elevated temperatures the fluoroborates
show an appreciable equilibrium pressure of gas-
eous BFs;. The equilibrium pressure®® above a

1000

900

3

fl
Q
o

TEMPERATURE (°C)
| (o))
O o
o O

3

300

 

200
NaoF 20 40 60 80
NaBF, (mole % )

NoBF4

Fig. 4. The system NaF-NaBF,.

FEBRUARY 1970
melt maintained at the eutectic composition (8
mole% NaF, 92 mole% NaBF4) is given by

5,920
U @)

 

log PTorr = 9.024 -

PHYSICAL PROPERTIES OF FUELS
AND COOLANTS

Tables II andIII list some of the pertinent phys-
ical properties® for MSRE and MSBR fuels and
secondary coolants.

Many of the properties shown are estimates
rather than measured values. These estimates
have been carefully prepared from the best avail-
able measurements on several salt mixtures of
similar composition. The values given are un-
likely to be in error sufficient to remove the fluid
from consideration. It is clear, however, from the
fact that estimates rather than measured values
are shown that an experimental program must be
devoted to firming up the physical properties of
these materials.

The densities were calculated from the molar
volumes of the pure components by assuming the
volumes to be additive. The heat capacities were
estimated by assuming that each gram atom in the
mixture contributes 8 calories per degree centi-
grade. The value of 8 is the approximate average
from a set of similar fluoride melts. The vis-
cosity of the MSBR fuel and the BeF:-based cool-
ants were estimated from other measured
LiF-BeF: and NaF-BeF: mixtures.

The vapor pressure of the fuels and the BeF'»
based coolants are considered negligible; extrapo-
lation of measurements from similar mixtures

TABLE II

Composition and Properties of MSRE
and MSBR Fuels

 

 

 

 

 

MSRE MSBR
Fuel Fuel
Composition LiF 65 LiF T1.7
(mole%) BeF, 29.1 | BeF: 16
ZrFs 5 ThF, 12
UF, 0.9 | UF,4 0.3
Liquidus
°C 434 500
°F 813 932
Properties at 600°C (1112°F)
Density, g/cm’ 2.27 3.35
Heat capacity, cal/(g °C)
or Btu/(lb °F) 0.47 0.33
Viscosity, cP 9 12
Vapor pressure, Torr Negligible | Negligible
Thermal conductivity,
W/(°C cm) 0.014 0.011

Grimes MSBR CHEMISTRY

 

 

TABLE III
Composition and Properties of Possible MSBR Secondary Coolants
C C. Cs
Composition (mole%) NaF 8 | LiF 23
NaBFs 92 | NaF 41 NaF 57
BeF, 36 BeF: 43
Liquidus Temperature:
°C 385 328 340
°F 725 622 644
Physical Properties al
850°F (454°C)*
Density, 1b/ft® 121 136 139
Heat capacity,
Btu/(1b °F) 0.36 0.47 0.44
Viscosity, cP 2.5 40 65
Vapor pressure at
1125°F (607°C),> mm 200°¢ Negligible | Negligible
Thermal conductivity,
W/(°C cm) 0.005 0.01 0.01

 

 

 

 

 

 

aMean temperature of coolant going to the primary heat exchanger.

b.'I-Iighest normal operating temperature of coolant.

cRepresents pressure of BF; in equilibrium with this melt compo-
sition.

yielded pressures < 0.1 mm. The partial pressure
of BFs above the fluoroborate coolant mixture was
calculated from measurements on very similar
mixtures.

CHEMICAL COMPATIBILITY OF MSBR MATERIALS

Details and specific findings of the large pro-
gram of corrosion testing are presented by McCoy
et a1? as a separate article in this issue. In brief,
compatibility of the MSBR materials is assured by
choosing as melt constituents only fluorides that,
insofar as is possible, are thermodynamically
stable toward the moderator graphite and toward
the structural metal, Hastelloy-N,® a nickel alloy
containing ~ 12% Mo, 7% Cr, 4% Fe, 1% Ti, and
small amounts of other elements. The major fuel
components (LiF, BeF2, UF4, and ThF4) are much
more stable than the structural metal fluorides
(NiF2z, FeF3z, and CrF2); accordingly, the fuel and
blanket have a minimal tendency to corrode the
metal. Such selection, combined with proper pur-
ification procedures, provides liquids whose cor-
rosivity 'is well within tolerable limits. The
chemical properties of the materials and the
nature of their several interactions are described
briefly in the following.

 

 

 

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

aHastelloy-N used in MSRE was Ni with 17% Mo, 7% Cr,
59 Fe. Probable composition of modified Hastelloy-N
for MSBE is that shown above.

FEBRUARY 1970 143
Grimes MSBR CHEMISTRY

Thermodynamic Data for Molten Fluorides

A continuing program of experimentation over
many years has been devoted to definition of the
thermodynamic properties of many species in
molten LiF-BeF: solutions. Many techniques have
been used in this study. Many of the data have
been obtained by direct measurement of equilib-
rium pressures for reactions such as

Hzg) + FeFyq)= Fe?c) + 2HF () (5)
and

2HF () + BeO(c)= BeF:(1) + H20(y) (6)

[where (g), (c), and (d) represent gas , crystalline
solid, and solute, respectively] using the molten
fluoride as reaction medium. Baes has reviewed
all these studies and by combining the data with the
work of others has tabulated thermodynamic data
for many species in molten 2LiF.BeF,.?® TableIV
below records pertinent data for the major com-
ponents of MSRE and MSBR fuels and for cor-
rosion products in molten 2LiF-BeF,.

From these data one can assess the extent to
which UFs-bearing melt will disproportionate ac-
cording to the reaction

4UFs(q) = 3UF4q) + U . | (7)

TABLE IV

Standard Free Energies of Formation for Species
in Molten 2LiF - BeFo

(773 to 1000°K)

 

 

 

-AG/ -AG1000°k)
Material? (kcal/mole) (kcal/mole)
LiF (1) 141.8-16.6 x 10~3T°K 125.2
BeFz (1) 243.9 - 30.0 x 10”3 7°K 106.9
UFs3 (q) 338.0-40.3 x 10 T°K 99.3
UF4(q) 445.9 - 57.9 x 102 T°K 97.0
ThFs (g 491.2-62.4 x 10 T°K 107.2
ZrFa(q) 453.0 - 65.1 x 103 7°K 97.0
NiFz (g 146.9 - 36.3 x 10> T°K 55.3
FeFa (g 154.7-21.8 x 1072 T°K 66.5
CrFz(q) 171.8-21.4 x 1073 T°K 75.2
MoFs (g) 370.9-69.6 x 10> T°K 50.2

 

 

 

 

*The standard state for LiF and BeF2 is the molten
2LiF - BeF2 liquid. That for MoFe (g) is the gas at one
atmosphere. That for all species with subscript (d) is
that hypothetical solution with the solute at unit mole
fraction and with the activity coefficient it would have
at infinite dilution.

144

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

For the case where the total uranium in the mol-
ten solution of 0.9 mole% (as in the MSRE) the
activity of metallic uranium is near 10~ '° with
1% of the UF, converted to UFs and is near
2 X 107" with 20% of the UF4 so converted.?” Op-
eration of the reactor with a small fraction
(usually <2%) of the uranium present as UF;s is
advantageous insofar as corrosion and the conse-
quences of fission are concerned (see subsequent
sections of this report). Such operation with some
UFs; present should result in formation of an
extremely dilute (and experimentally undetectable)
alloy of uranium with the surface of the container
metal. Operation with > 50% of the uranium as
UFs would lead to much more concentrated (and
highly deleterious) alloying and to formation of
uranium carbides. Fortunately, all evidence to
date demonstrates that operation with relatively
little UFs is completely satisfactory.

Oxidation (Corrosion) of Metals

The data of Table IV reveal clearly that in
reactions with structural metals (M)

2UF4 ) + M()= 2UF3(4) + MF2(q) (8)
Cr is much more readily attacked than is iron,
nickel, or molybdenum.?” Few thermodynamic
data exist for the fluorides of titanium in the pure
state and none for dilute solutions in molten
fluoride solvents. Estimates®® of free energies of
formation suggest that Ti is somewhat more reac-
tive than Cr. Titanium on the surface layers of
the metal should, therefore, be expected to react
with the available oxidants, such as UF,. Such
oxidation and selective attack follow from reac-
tions such as the following:
Impurities in the melt

Cr + NiF, - CrF, + Ni (9)
or
Cr + 2HF — CrF, + H, (10)
Oxide films on the metal
NiO + BeFz2 — NiF: + BeO , (11)
followed by reaction of NiF, with Cr.
Reduction of UF4 to UF;
Cr + 2UF4 = 2UF; + CrF» (12)

Reactions (9), (10), and (11) above will proceed
essentially to completion at all temperatures
within the MSBR circuit. Accordingly, such reac-
tions can lead (if the system is poorly cleaned) to
a noticeable, rapid initial corrosion rate. How-
ever, these reactions do not give a sustained
corrosive attack.

FEBRUARY 1970
The reaction of UF4 with Cr, on the other hand,
has an equilibrium constant with a small tempera-
ture dependence; hence, when the salt is forced to
circulate through a temperature gradient, a pos-
sible mechanism exists for mass transfer and
continued attack.

If nickel, iron, and molybdenum are assumed to
be completely inert diluents for chromium (as is
approximately true), and if the circulation rate in
the MSBR is very rapid, the corrosion process can
be simply described. At high flow rates, uniform
concentrations of UFs; and CrF. are maintained
throughout the fluid circuit; these concentrations
satisfy (at some intermediate temperature) the
equilibrium constant for the reaction. Under these
steady-state conditions, there exists some tem-
perature (intermediate between the maximum and
minimum temperatures of the circuit) at which the
initial surface composition of the structural metal
is at equilibrium with the fused salt. Since the
equilibrium constant for the chemical reaction in-
creases with increasing temperature, the chromi-
um concentration in the alloy surface tends to
decrease at temperatures higher than T and tends
to increase at temperatures lower than 7. [In
some melts (NaF-LiF-KF-UF4, for example) AG
for the mass transfer reaction is quite large, and
the equilibrium constant changes sufficiently as a
function of temperature to cause formation of
dendritic chromium crystals in the cold zone.]
For MSBR fuel and other LiF-BeF2-UF4 mixtures,
the temperature dependence of the mass-transfer
reaction is small, and the equilibrium is satisfied
at reactor temperature conditions without the for-
mation of crystalline chromium. Thus, in the
MSBR, the rate of chromium removal from the
salt stream by deposition at cold-fluid regions is
controlled by the slow rate at which chromium
diffuses into the cold-fluid wall; the chromium
concentration gradient tends to be small, and the
resulting corrosion is well within tolerable limits.
Titanium, which must be presumed to undergo
similar reactions, diffuses less readily than does
Cr in Hastelloy-N. It seems most unlikely that the
1% of Ti in the alloy will prove detrimental.

The results of numerous long-term tests have
shown that Hastelloy-N has excellent corrosion
resistance to molten-fluoride mixtures at temper-
atures well above those anticipated in MSBR.” The
attack from mixtures similar to the MSBR fuel at
hot-zone temperatures as high as 1300°F is barely
observable in tests of as long as 12 000 h. A fig-
ure of <0.5 mil/year is expected.’® In MSRE,
attack was ~ 0.1 mil/year. Thus, corrosion of the
container metal by the reactor fuel does not seem
to be an important problem in the MSBR.™

It is likely that the NaF-NaBF4 coolant mixture
will also prove compatible with Hastelloy-N, but

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

FEBRUARY 1970

Grimes MSBR CHEMISTRY
experimental study of this is much less extensive.
The free energy change for the chemical reaction

BFs(p) + 3/2Cr (s— 3/2CrFzq) + B(s) (13)

is ~+30 kcal at 800°K.” The reaction is, there-
fore, quite unlikely to occur, and similar reactions
with Fe, Mo, and Ni are much less so. In addition,
the above reaction becomes even less likely (per-
haps by 5 kcal or so) when one considers the
energetics of formation of the compound NaBF,
and dilution of the NaBF4 by NaF. However, re-
actions which produce metal fluorides and metal
borides are those which must be anticipated. Ti-
tanium, a minor constituent of Hastelloy-N, has a
quite stable boride.’® The reaction

2BFs(p) + 2Ti(o= TiB2(c)+ 2TiFsa  (14)
probably shows a small negative free energy
change and must be expected to proceed. The very
small diffusion rate of Ti in Hastelloy-N would be
expected to markedly limit the reaction. Thermo-
chemical data for the borides of Fe, Cr, Ni, and
Mo do not seem to have been established. Unless
the borides of these materials are very stable
(AGf values more negative than -25 kcal B atom)
the Hastelloy-N should prove resistant to pure
NaF-NaBF4 coolant.

Compatibility of Graphite with Fluorides

Graphite does not react with molten fluoride
mixtures of the type to be used in the MSBR.
Available thermodynamic data®® suggest that the

most likely reaction

4UF4 + C = CF4 + 4UF; (15)

should come to equilibrium at CF4 pressures
£ 10°% atm. CF4 concentrations over graphite-salt
systems maintained for long periods at elevated
temperatures have been shown to be below the
limit of detection (< 1 ppm) of this compound by
mass spectrometry. Moreover, graphite has been
used as a container material for many NaF-ZrF,-
UF., LiF-BeF:-UF4, and other salt mixtures with
no evidence of chemical instability.

CHEMICAL BEHAVIOR IN MSRE

The Molten-Salt Reactor Experiment was, as
detailed in an article by Haubenreich and Engel™
operated during much of 1966, 1967, and early
1968 with the original fuel charge in which the
uranium was 33% 2°°U with the balance consisting
primarily of >**U. The reactor accumulated nearly
70 000 MWh of operation during this interval with
the major fraction of this accumulated at the max-
imum power of ~ 8 MW(th).

145
Grimes MSBR CHEMISTRY

Behavior of Fuel Components

Samples of the molten salts were removed
routinely from the fuel and coolant circuits and
were analyzed for uranium, major fuel constitu-
ents, possible corrosion products, and (less fre-
quently) for oxide ion contamination. Early in the
power runs, standard samples of fuel were drawn
three times per week; this schedule was markedly
diminished as confidence in the system grew. At
late stages in the power runs, analyses for these
items was generally done on a one-per-week
basis. The LiF-BeF: coolant system was sampled
every two weeks initially and less frequently dur-
ing 1967 and 1968.

Chemical determinations of uranium content of
the fuel salt were run by coulometric titrations® of
dilute aqueous solutions. Such analyses showed
good reproducibility and high precision (+0.5%).
(On-site reactivity balance calculations proved to
be about ten-fold more sensitive than this in es-
tablishing changes in uranium concentrations with-
in the fuel circuit.) Agreement of the chemically
determined uranium concentrations (in a statis-
tical sense) with the reactor inventory values
diminished by the uranium burnup during opera-
tion was excellent. The inventory value at end of
operation on °*°U showed the uranium concentra-
tion to be 4.532% by weight.*®* The difference (200
g of uranium out of 220 kg) is <0.1%. These data
strongly indicate that uranium losses (as to the
purge gas stream) have been extremely small and
that, as anticipated, the fuel salt is chemically
stable during long periods of reactor operation. A
further check on these numbers will be possible
when the **°U-?°*U mixture removed by fluori-
nation from the MSRE fuel is recovered and
assayed.

While analyses for UFs and for ZrF, agree
quite well with the inventory data, the values for
"LiF and BeF: have never done so; analyses for
LiF have shown higher and those for BeF, have
shown lower values than the inventory since
start-up. Table V shows a comparison of typical
analysis with the original inventory: value. This

TABLE V

Typical and Original Composition
of MSRE Fuel Mixture

 

 

Original Value
Constituent (mole%) Typical Analysis
"LiF 63.40 + 0.49 64.35
BeFs 30.63 + 0.55 29.83
ZrF4 5.14 £ 0.12 5.02
UF4 0.821 + 0.008 0.803

 

 

 

 

 

146 NUCLEAR APPLICATIONS & TECHNOLOGY

discrepancy in LiF and BeF: concentration re-
mains a puzzle, but there was nothing in the an-
alyses (or in the behavior of the reactor) to
suggest that any changes occurred.

Oxide concentration in the radioactive fuel salt
were performed by careful evaluation of H: O pro-
duced upon treatment of the salt samples with
anhydrous gaseous HF. All samples examined
showed less than 100 ppm of O%; no perceptible
increase in the values with time or with opera-
tional details was apparent.’* These facts are
quite reassuring insofar as maintaining oxide
contamination at very low levels in future reactor
systems is concerned.

MSRE maintenance operations have necessi-
tated flushing the interior of the drained reactor
circuit on numerous occasions. The salt used for
this operation consisted originally of a "LiF-BeF,
(66.0 to 34.0 mole%) mixture. Analysis of this salt
before and after each use shows that ~ 215 ppm of
uranium is added to the flush salt in each flushing
operation; this corresponds to removal of 22.7 kg
of fuel-salt residue (~0.5% of the charge) from
the reactor circuit.

Behavior of Corrosion Products

Analysis of many samples at regular intervals
during reactor operation showed relatively high
values for iron (120 ppm) and nickel (50 ppm) in
the salt.”””™° These values scattered significantly
but showed no perceptible trends. They do not
seem to represent dissolved Fe®* or Ni**. Molyb-
denum concentrations were shown, in the few
attempts made, to be below the detectible limit
(~ 25 ppm) for chemical analysis.

The chromium concentration, as determined by
chemical analysis, rose from an initial value near
40 ppm to a final value after 70 000 MWh of
~85 ppm. A considerable scatter (+ 15 ppm) in
the numbers was apparent but the increase with
time and reactor operation is clearly real. All
evidence suggests that the analytically determined
chromium was largely, if not entirely, present as
dissolved Cr**. This observed increase in chro-
mium concentration corresponds to removal of
<250 g of this element from the MSRE circuit. If
this were removed uniformly it would deplete the
chromium in the ailoy to a depth of < 0.2 mil.
Such an estimate of corrosion seems consistent
with observations (reported elsewhere in this
series) showing very slight attack by the fuel
during MSRE operation.

This virtual absence of corrosion—though in
general accord with results from a variety of en-
gineering corrosion tests—is mildly surprising
since the MSRE was operated for appreciable
periods with the fuel salt more oxidizing than

VOL. 8 FEBRUARY 1970
necessary. The analysis for ratio of U*t to U
has proved especially difficult in the radioactive
salts, but it seems certain that the MSRE fuel has,
at times, contained < 0.2% of the total uranium as

U

The fission process (see following section of

this report) appears to be mildly oxidizing toward
dissolved U°t in the fuel. Accordingly, and al-
though the corrosion product analysis did not
seem to require it, a convenient means was de-
veloped for adding U°Y to the reactor circuit as
desired. Beryllium metal (as 3-in. rods of 3-in.
diameter encased in a perforated capsule of
nickel) was introduced through the sampling sys-
tem and suspended in the flowing salt stream in
the MSRE pump bowl. This active metal reacts
with UF4 by

Be(()c) + 2UF4q)— BeFza)+ 2UF3(q) (16)
at a rate such that some 600 g of UFs is produced
during an 8-h treatment. The salt flowing past the
Be appears to be locally over-reduced so that at
least 6% of the UF4 becomes UFjs; crystalline Cr
was observed upon the nickel capsule from one
such reduction attempt. The over-reduced salt
mixture clearly reacted and achieved equilibrium
with the large excess of unreduced salt in the
pump bowl and reactor circuit.

A molten-salt reactor which included a repro-
cessing circuit could use that external circuit to
maintain the UF,/UFs; ratio at the desired level.
Use of techniques such as the Be addition for such
a reactor would seem to be quite feasible but
unnecessary.

Behavior of Fission Products

When fission of UF4 occurs in a well-mixed
high-temperature molten-fluoride, four fluoride
jons associated with the U** are released and the
two fission fragments must come quickly to a
steady state as common chemical entities. This
steady state is made very complicated by the
radioactive decay of many species. The valence
states that the fission product assume in the mol-
ten system are, presumably, defined by the re-
quirements that cation-anion equivalence be
maintained in the liquid and redox equilibria be
established both among the components of the melt
and the surface layers of the container metal.®”®
Fluorine and uranium species higher than 4 are,
in MSRE and MSBR fuels, unstable toward reduc-
tion by the active constituents of Hastelloy-N; the
fission product cations must satisfy the four
fluoride ions plus the fission product anions.
Should they prove inadequate, or if they become
adequate only by assuming oxidation states incom-
patible with Hastelloy-N, then oxidation of UFs

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

Grimes MSBR CHEMISTRY
must occur or the container must supply the de-
ficiency. Operation of MSRE has indicated that the
presence of UF; in relatively low concentrations
(see preceding sections) suffices to avoid oxida-
tion or corrosion from this source.

Behavior of the fission products in MSRE was
studied®®*™*° during the entire period of operation
with 23°U. Samples of MSRE fuel, drawn from the
sampling station in the MSRE pump bowl, have
been routinely analyzed for numerous fission prod-
uct species. Since samples drawn in open metal
cups were shown (for reasons described below) to
give erroneously high values for gas-borne spe-
cies, evacuated bulbs of Ni sealed with a fusible
plug of salt were used to obtain many of the data
presented here.

Samples of the helium gas purge within the
MSRE were obtained by use of identical evacuated
sample bulbs in which the fusible plug was per-
mitted to melt with the bulb exposed to the gas
phase.*"»%®

In addition to these salt and gas samples, as-
semblies of surveillance specimens were exposed
as the central stringer within the MSRE core.
These assemblies included specimens of various
graphites and many specimens of Hastelloy-N and
some other metals. Their removal at reactor
shutdowns permitted study of fission product de-
positions after 8000, 20 000, and 70 000 MWh of
operation.

Samples of salt removed from the pump bowl
were dissolved in analytical chemistry hot cells
(after careful leaching of activities from the out-
side of capsules where necessary); the fission
products were identified and their quantity de-
termined radiochemically. Analysis of the gas
samples was accomplished similarly after careful
leaching of the activities from within the capsules.
Deposition of fission products upon the metallic
surveillance specimens was determined in the
same way after repeated leachings of the metal
surface.

Deposition on the graphite specimens was es-
tablished as a function of depth of penetration.
This was accomplished by careful milling of suc-
cessive thin layers from each surface of the rec-
tangular specimens. The removed layers were
separately recovered and analyzed for several
fission product isotopes.’®™*°

The fission gases Kr and Xe form no com-
pounds under conditions existing in a molten-salt
reactor,*’ and these elements are only very spar-
ingly soluble in the molten fluoride.*” ** The
helium purge gas introduced into the MSRE pump
bowl serves to strip these gases from the in-
coming salt to charcoal filled traps downstream in
the exit gas system. This stripping ensures that
the fission product daughters of these gases will

FEBRUARY 1970 147
Grimes MSBR CHEMISTRY

appear within the reactor system at lower than the
theoretical concentration. The MSRE graphite is,
of course, permeable to Kr and Xe; radioactive
daughters of these gases are, accordingly, ex-
pected in the graphite specimens.

The fission products Rb, Cs, Sr, Ba, the lan-
thanides and yttrium, and Zr all form quite stable
and well-known fluorides which are soluble in
MSRE fuel.*>*® These fission products were ex-
pected (except to the extent that volatile precursor
Kr and Xe had escaped) to be found almost com-
pletely in the molten fuel. These expectations
have been confirmed. Isotopes such as °°Zr, °'Sr,
and "*Cewhich have no precursors of consequence
are typically found in the salt at 90%% of the cal-
culated quantity. Isotopes such as ®°Sr and “°Ba
whose Kr and Xe precursors have appreciable
half-lives are found in less than the calculated
quantity; these isotopes are, as expected, found in
the graphite specimens and the gas samples as
well,

The elements (molybdenum, niobium, ruthen-
ium, and tellurium) whose higher valence fluorides
are volatile and relatively unstable toward re-
duction by UF;%'%2%%%%7 are virtually absent from
the salt. Typical analyses of samples (which have
been carefully protected from the pump bowl gas
by use of the evacuated bulb samples) show
< 2% of the calculated concentration of these iso-
topes.

Deposition of fission products on the ~ 7- to
10-mil outer layers of graphite specimens ex-
posed for 8000 MWh in MSRE is shown in Table VL
It is clear that, with the assumption of uniform
deposition in or on the moderator graphite, ap-
preciable fractions of Mo, Te, and Ru and a larger
fraction of the Nb are associated with the graph-
ite. Subsequent examinations of graphite after
22 000 and 70 000 MWh have shown somewhat
smaller fractions deposited for all these isotopes.
The data of Table VII described below show values
more typical of the longer exposures.*°

A careful determination of uranium in or on the
graphite specimens led to values < 1 pug/cm?.%
This quantity of uranium—equivalent to a few
grams in the moderator stack—seems completely
negligible.

Figure 5 shows the change in concentration of
the fission product isotopes with depth in the
graphite. Those isotopes (such as ™°Ba) which
penetrated the graphite as noble gases show
straight lines on the logarithmic plot; they seem
to have remained at the point where the noble gas
decayed As expected, the gradient for “°Ba with
16- sec **Xe precursor is much steeper than that
for **Sr, which has a 3.2-min **Kr precursor. All
the others shown show a much steeper concentra-
tion dependence. Generally the concentration

148 NUCLEAR APPLICATIONS & TECHNOLOGY

TABLE VI

Fission Product Deposition on Surface*
of MSRE Graphite

 

 

 

Graphite Location
Top Middle Bottom
Percent of Percent of Percent of
Isotope Total? Total? Total?
PMo 13.4 17.2 11.5
1%27e 13.8 13.6 12.0
193pu 11.4 10.3 6.3
*Nb 12 59.2 62.4
131y 0.16 0.33 0.25
Bz 0.33 0.27 0.15
%oe 0.052 0.27 0.14
®sr 3.24 3.30 2.74
140 1.38 1.85 1.14
Hloe 0.19 0.63 0.36
137cs 0.07 0.25 0.212

 

 

 

 

 

 

*Average of values of 7- to 10-mil cuts from each of
three exposed graphite faces.

aPercent of total 1n reactor deposited on graphite if each
cm® of the 2 x 10° cm? of moderator had the same con-
centration as the specimen.

drops a factor of 100 from the top 6 to 10 mils to
the second layer. More recent examinations which
used cuts as thin as 1 mil show the materials to
be concentrated in an extremely thin layer. Typ-
ically 95% of the metallic species are within
3 mils of the surface.

 

 

 

 

TABLE VI
Fission Product Distribution for MSRE at End of **®*U Operation
Quantity Found (% of Calculated Inventory)
On In
On Hastelloy- | Purge
Nuclide In Fuel Graphite? N2 Gas? | Total
Mo 0.17 9.0 19 50 78
132e 0.47 5.1 9.5 74 89
129ne 0.40 5.6 11.5 31 48
193 pu 0.033 3.5 2.5 49 55
1%%pu 0.10 4.3 3.2 130 138
**Nb 0.001 to 2.2 41 12 11 64
Bzr 94 0.14 0.06 0.43 95
8%sr 83 8.5 0.08 17 109
1op, 96 1.9 0.05 0.48 98
Hloe 0.33 0.03 0.88
e 0.92 0.03 2.7
131y 60 0.11 0.3 19 80

 

 

 

 

 

 

 

2Calculated on assumption that average values for surveillance
specimens are representative of all graphite and metal surface in
MSRE.

PThese values are the percentage of production lost assuming mean
values found in gas samples apply to the 4-liter/min purge.

VOL. 8 FEBRUARY 1970
DISINTEGRATIONS PER MINUTE PER GRAM OF GRAPHITE

 

0 10 20 30 40 50
DISTANCE FROM SURFACE OF GRAPHITE (mils)

Fig. 5. Concentration profile of fission products in

MSRE core graphite after 8000 MWh,

Examination of the Hastelloy-N surveillance
specimens exposed in the MSRE core reveals very
low concentrations of the stable fluoride-forming
fission products such as *8r, ®*Zr, and *'Ce. The
more noble metals such as Mo, Nb, and Ru are
deposited, along with Te, in appreciable though not
in large quantities. If the surveillance specimens
are truly representative of all the metal surface
in MSRE then (see Table VII) only a relatively
small fraction of these metals are so deposited.
However, flow conditions in the MSRE heat ex-
changer (where most of the metal surface exists)
is turbulent while laminar flow occurs over the
core specimens. It seems very likely that the
actual fraction deposited on MSRE metal is higher
than that shown here.

Analysis of samples of gas from MSRE yielded
some real surprises. Isotopes such as 8gr and
0By are found in amounts consistent with con-
sideration of the half-lives of their Kr and Xe

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

Grimes MSBR CHEMISTRY
precursors. A small quantity of *’Zr appears in
the gas phase; this isotope, along with several
others with stable soluble fluorides, almost cer-
tainly appears in the gas as fine particles of salt
mist.>® However, high concentrations of “°Mo,
1327¢  ®Nb, ©Ru, and Ru are found in the gas
phase. Indeed, as Table VII indicates, the gas
samples represent the largest fractions observed
for each of these isotopes except *°Nb. In addition,
appreciable fractions of '°Ag and of palladium
isotopes have been observed in the purge gas
samples. An appreciable quantity of '°'I appears
in the gas phase; it seems very likely that its
appearance there is due to ‘‘volatilization’’ of its
tellurium precursor.

The mechanisms by which the Mo, Nb, Te, and
Ru isotopes appear in the gas phase are still not
fully understood. NbFs, MoFs, RuFs, and TeFq are
all volatile (although Ag and Pd have no volatile
fluorides); these volatile fluorides, however, are
far too unstable, with the possible exception of
NbFs, to exist at the redox potentials in MSRE.
Stability of lower fluorides of these elements is
poorly known; it is possible that some of these
may contribute to the ‘‘volatility.”’

It seems much more likely, however, that all
these species exist in the reduced MSRE fuel in
elemental form. They originate as (or very rap-
idly become) individual metal atoms. They ag-
gregate in the fuel at some slow but finite rate,
and become insoluble as very minute colloidal
particles which then grow at a slower rate. These
colloidal particles are not wetted by the fuel, they
tend to collect as gas-liquid interfaces, and they
can readily be swept into the gas stream of the
helium purge of the pump bowl. They tend to plate
upon the metal surfaces of the system, and (as
extraordinarily fine ‘‘smoke’’) to penetrate the
outer layers of the moderator. While there are
difficulties with this interpretation it seems more
plausible than others suggested to date.

Deposition of fission products on or in the
graphite is, of course, undesirable since there
they serve most efficiently as nuclear poisons. Of
the species studied only Nb (whose carbide is

. stable at MSRE temperatures) seems to prefer the

graphite.

It seems likely that in a MSBR the molten fuel
will contain virtually all of the zirconium, the
lanthanides and yttrium produced, and a large
fraction of the iodine, rubidium, cesium, stron-
tium, and barium. Thess species would, therefore,
be available for removal in the processing plant.
The salt will contain very little Mo, Nb, Te, Ru,
and (presumably) technetium. These will report
to the MSBR metal, graphite, and purge gas sys-
tems in fractions which are probably not very
different from those indicated for MSRE.

FEBRUARY 1970 149
Grimes MSBR CHEMISTRY

SEPARATIONS CHEMISTRY

Molten-salt breeder reactors will require ef-
fective schemes for recovery of the fissionable
material which is produced and for removal of
fission product poisons from the fuel. MSRE pro-
duced no significant quantity of fissionable isotope
and made no provision for on-stream removal of
fission products other than the noble gases Kr and
Xe.

Recovery of uranium from molten fluorides by
volatilization as UFs and the subsequent decon-
tamination of this UFe by sorption-desorption on
beds of NaF is well demonstrated. Recovery of
uranium from MSRE or MSBR fuel by this scheme
is clearly feasible. However, performance of a
single-fluid MSBR is markedly improved if the
bred protactinium is removed from the circulating
fuel and permitted to decay to 2**U outside the
reactor core. Fluorination is not effective in re-
moval of protactinium from molten fluoride mix-
tures. Development of other processes for Pa
recovery has, accordingly, been necessary.

In molten-salt reactors, from which Kr and Xe
are effectively removed, the most important fis-
sion product poisons are among the lanthanides.
These fission products, as indicated above, form
stable fluorides which are soluble in the fuel and
will report almost quantitatively with the fuel to
the separations plant. Their removal especially
from fuels with high ThFs concentrations is diffi-
cult. Rare earths have, accordingly, received
more attention than other fission products.

Processes for recovery of Pa and the lan-
thanide fission products from MSBR fuels are
described in some detail by Whatley et al.*® The
chemical basis for such processes is described
very briefly below.

Separation of Protactinium

Protactinium is produced in the MSBR fuel by
reaction of a neutron with **?*Th to yield 2**Th
which transmutes to ***Pa; the ***Pa decays with a
27.4-day half-life to °**U. Clearly, a separation
of **Pa from the fuel needs, if it is to be effec-
tive, to be accomplished on a cycle time short
compared with this half-life. It is desirable to
process the fuel on a cycle of three to five days.
The process needs, therefore, to be relatively
simple.

Removal of Pa from LiF-BeF;-ThF,; mixtures
by addition of BeO, ThO., or even UO: has been
demonstrated in laboratory-scale experiments?®®
Addition of these oxides to MSBR fuel composition
would, as indicated in a previous section of this
report, result in a uranium-oxide-rich solid solu-
tion of UOz-ThO2. Whether the protactinium oxide

150 NUCLEAR APPLICATIONS & TECHNOLOGY

was adsorbed on the added oxide or participated in
the solid solution formation is not known. The
process was shown to be reversible; treatment of
the oxide-fluoride system with anhydrous HF dis-
solves the oxide and returns the Pa to the fluoride
melt from which it can readily be reprecipitated.
It seems likely that Pa might be removed from
the MSBR fuel by effective oontact of a fuel mix-
ture side stream with UO»-ThO, solid solution.
Such a possibility—which is still under considera-
tion in laboratory-scale equipment—would add
appreciably to the **U inventory through holdup of
UO: in the solid solution. It would also probably
require that the oxide ion concentration in the fuel
after contact be diminished (as by treatment with
HF) before the melt could be returned to the
reactor.

Separations based upon reduction of the Pa
(present in the fuel as PaF,) to metal and extrac-
tion into liquid metals appears at present to be
more attractive.*® Such separations have the fol-
lowing basis.

When thorium metal, dissolved as a dilute alloy
in bismuth (Thg;), is equilibrated with a molten
mixture of LiF, BeF:, and ThF, several equilibria
must be satisfied simultaneously. These include

LiF (4) + 1 Th(gj) = i ThFq4(q) + Li(g;) (17)
BeF2(d4)+ 3Th(pi) =3ThFs + Be(gj (18)
1 ThF4() + i Th(gy = ThFs g , (19)

where (d) and (Bi) indicate that the species are in
fluoride solution and alloyed with the Bi, respec-
tively. Reaction (19) seems to be of little im-
portance since only a small quantity (if any) ThFs;
forms. Reaction (18) does not occur perceptibly;
Be is “‘insoluble’’ in Biand forms no intermetallic
compounds with that liquid metal. Reaction (17)
occurs to an appreciable extent even though LiF
(see Table IV) is much more stable than ThF,.'°

Uranium forms stable dilute alloys in Bi, and
its fluorides are less stable than ThF4.' Accord-
ingly, the reactions

UF4(q)+ sTh(pi) = 1 ThF4(q) + UFs(q) (20)

and

UFs(q) + 1 Thgi)= U(gi) + 3 ThF4g) (21)
proceed very far to the right with low concentra-
tions of Th (and Li) in the Biat equilibrium. Uran-
ium 1is, therefore, very readily extracted from
MSBR fuel melts. *'%2

It the LiF-BeF:-ThF, melt contains PaF,, the
reaction

PaF4(d) + Th(Bi) - ThF4(d) + Pa(Bi) (22)

VOL. 8 FEBRUARY 1970
proceeds to quite a convenient extent®® "*°as shown

in Fig. 6 although protactinium is reduced less
easily than is uranium. Accordingly, if a melt
containing both UFs and PaF,4 is contacted with
bismuth and thorium metal added in small incre-
ments the uranium and the protactinium are ex-
tracted.

Preliminary experiments® disclosed that Pa
could indeed be removed from the salts by reduc-
tion with Th but material balance data were very
poor and little Pa appeared in the Bi. More recent
experiments have used equipment of molybdenum,
which seems essentially inert to the salt and the
alloy, and very careful purification of the ma-
terials and the cover gas.*” > By these techniques
two sets of investigators®’™ have shown values in
excess of 1500 for the separation factor

= —_— (23)
Th  Npar, NTh

a
where N indicates mole fraction of the indicated
species either in metal or in LiF-BeF2:-ThF4 so-
lution. The separation factor for U from Pa has
been shown to be above 25. These separation fac-
tors are, of course, quite adequate for process
design, as discussed in detail by Whatley et al.*®

 

100§ VN
|

90

 

 

80 ¢4 a o

 

 

 

 

 

 

o PROTACTINIUM IN BISMUTH
o PROTACTINIUM IN SALT
s THORIUM METAL IN BISMUTH

70

 

 

0 |

 

 

 

FLOWING ARGON
STATIC ARGON

 

 

 

50 “ FLOWING -

. FLOWING ARGON A HELIUM
]

 

o

|

|

 

40

 

 

 

30

 

20

Z 4t

o

>

 

I

THORIUM METAL ADDED OR TOTAL PROTACTINIUM (percent)

 

 

 

 

 

 

 

L
l |
I I
| I
I
| I
| I
| |
|
I
A
I
|
|
I.
I

 

 

 

 

 

|
|
I
I
l
I
|
|
A A | a A
A
|
|
|
I
I

 

 

CONTACT TIME (h)

Fig. 6. Distribution of protactinium and material bal-
ance for equilibrium of LiF-BeF,-ThF, (72-16-

12 mole%) with Bi-Th alloy (0.93 wt% Th).

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

b L
0 S 25 45 65 85 105 125 145 165

Grimes MSBR CHEMISTRY

Separation of Rare Earths

There is no real doubt that the rare-earth
elements, which form fluorides that are more
stable than ThF.' will be present in MSBR fuel
sent to the processing plant.

The limited solubility of these trifluorides
(though sufficient to prevent their precipitation
under normal MSBR conditions) has suggested a
possible recovery scheme. When a LiF-BeF:-
ThF,-UF. melt (in the MSBR concentration range)
that is saturated with a single rare-earth fluoride
(LaFs, for example) is cooled slowly, the precipi-
tate is the pure simple trifluoride. When the melt
contains more than one rare-earth fluoride the
precipitate is a (nearly ideal) solid solution of the
trifluorides.*®*® Accordingly, addition of an ex-
cess of CeFs; or LaFs to the melt followed by
heating to effect dissolution of the added tri-
fluoride and cooling to effect crystallization ef-
fectively removes the fission product rare earths
from solution. It is likely that effective removal
of the rare earths and yttrium (along with UFs and
PuF;) could be obtained by passage of the fuel
through a heated bed of solid CeFs or LaFs. How-
ever, the price is almost certainly too high; the
resulting fuel solution is saturated with the sca-
venger fluoride (LaFs or CeFs, whose cross sec-
tion is far from negligible) at the temperature of
contact.

Similar solid solutions are formed by the rare-
earth trifluorides with UFs. It might, accordingly,
be possible to send a side-stream (with the **Pa
and 2**U removed by methods described above)
through a bed of UF; to remove these fission
product poisons; 2**UF; would presumably be used
for economic reasons. The resulting LiF-BeF:-
ThF, solution would be saturated with UFs after
its passage through the bed. This ***U would have
to be removed (for example, by electrolytic re-
duction®® into molten Bi or Pb) before the salt
could be returned to the cycle. The UF; bed could
be recovered by fluorination to separate the ura-
nium and rare earths. While this process probably
deserves further study, the instability of UF; in
melts with high UFs/UF, ratios and the ease with
which uranium alloys with most structural metals
would tend to make application of such a process
unattractive.

Removal of rare-earth ions (and other ionic
fission-product species) by use of cation ex-
changers has always seemed an appealing possi-
bility. The ion exchanger would, of course, need
(a) to be quite insoluble, (b) to be extremely un-
reactive (in a gross sense) with the melt, and
(c) to take up rare-earth cations in exchange for
ions of low-neutron cross section. The bed of
CeF; described above functions as an ion ex-

FEBRUARY 1970 151
Grimes MSBR CHEMISTRY
changer; it fails to be truly beneficial because it
is too soluble in the melt.

Unfortunately, not many materials are truly
stable to the LiF-BeF:-ThF4.-UF, fuel mixture.
Zirconium oxide is stable (in its low temperature
form) to melts whose Zr*'/U*" ratio is in excess
of ~3, and UO:-ThO: solid solutions are stable at
equilibrium U/Th ratios. It is conceivable that
sufficiently dilute solid solutions of Ce, O; in
these oxides would be stable and would exchange
Ce’" for other rare earth species. Intermetallic
compounds of rare earths with moderately noble
metals (or rare earths in very dilute alloys with
such metals) seem unlikely to be of use because
they are unlikely to be stable toward oxidation by
UF4. Compounds with oxygenated anions (such as
silicates and molybdates) are decomposed by the
fluoride melt; they precipitate UO., from the fuel
mixture. It is possible that refractory compounds
(such as carbides or nitrides) of the rare earths,
either alone or in solid dilute solution with an-
alogous uranium compounds, may prove useful.

By Reduction

The rare-earth fluorides are very stable
toward reduction to the metal. For example, at
1000°K the reaction

2/3LaFs () + Be()— 2/3La(s+ BeFzqy , (24)
where (c) and (1) indicate crystalline solid and liq-
uid, respectively, shows ~+30 kcal for the free en-
ergy of reaction.'® With the LaF; in dilute solution
and BeF; in concentrated solution in LiF-BeF, mix-
ture the situation is, of course, even more un-
favorable. However, the rare-earth metals form
extremely stable solutions®™ in molten metals such
as bismuth; the activity coefficient for La at low
concentrations in Bi is near 10™%, Therefore, the
reaction

ZLaF; @) +Be) = %Lapj + BeFz2) ,  (25)
where (d) indicates that the species is dissolved in
2LiF*BeF2, and (c) indicates crystalline solid can
be made to proceed essentially to completion.
Accordingly, LaF; can be reduced and extracted
into molten Bi from LiF-BeF. mixtures. Figure 7
shows the behavior of several rare-earth ele-
ments™ upon extraction into bismuth using Li°
(which is more convenient to use) as the reductant.
Such a process would seem to be useful,

However, when the melt is complicated, as is
necessary for the single-fluid breeder, by the
addition of large quantities of ThF, the situation
becomes considerably less favorable. Use of Li°
concentrations as high as those shown in Fig. 7 is
prohibited since the reaction

152 NUCLEAR APPLICATIONS & TECHNOLOGY

o
o

NEODYMIUM

o
N

LANTHANU

o

o
o

O

SAMARIUM

DISTRIBUTION OF RARE EARTH (mole fraction in metal / mole fraction in salt)

1072

 

1072 107! 10° To}
LITHIUM FOUND IN METAL PHASE (at.%)

Fig. 7. Effect of lithium concentration in metal phase

on the distribution of rare earths between LiF-
BeF, (66-34 mole%) and bismuth at 600°C,

2Bi(j) + $ThF4)+ Li(pi) = LiF(q)+ $ThBiz(c) (26)

proceeds at lower concentrations than these to
yield solid ThBi:. Accordingly, the most reducing
metallic phase which is tolerable is that repre-
sented by the (barely) saturated solutlon of ThBi,
in Bi. When such solutions are used the rare-
earth fluoride is much less completely extracted
and relatively small separation factors from tho- .
rium are obtained.®™:*® Figure 8 shows data
obtained® in extraction of Ce at 600°C from sev-
eral LiF-BeF:-ThF, mixtures (the points are
labeled to indicate mole percentages of LiF, BeF;,
and ThFy4, respectively) in this way. Separation
factors appreciably above unity are obtained even
in the worst case, and compositions containing
12 mole% ThFs with < 20 mole% BeFs; appear
somewhat more attractive. Unfortunately, Ce’+ is
one of the easiest of the rare earths to reduce.

VOL. 8 FEBRUARY 1970
Others in the series show even smaller separation
factors. Another paper in this series presents in-
formation showing how processes with separation
factors of this magnitude could be operated.*®
There is little doubt, however, that improved
(larger) separation factors would be desirable;
there is reason to believe that improvement can
be obtained by modification of the alloy phase.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6 T
74-20-6
5
. ®
_ T
z 9 / ®
0 | |
Na .
[ }]
& :
8 i
« Y
O N o
G3l——ti—FR
3 RS
S
o
< © //
g » ° / NOTE:
aQ / VALUES CALCULATED
n . AT THORIUM SATURATION
3 IN BISMUTH
1
M
{ o
0
w0
O
-20 -10 0 10 20 30

EXCESS FREE FLUORIDE
= mole % LiF —2(mole % Ber) -3(mole % ThFy,)

Fig. 8. Effect of salt composition on the separation of
cerium from thorium during reductive extrac-
tion of cerium from LiF-BeF,-ThF, mixtures
into bismuth at 600°C.

SUMMARY

Phase behavior of LiF-BeF:2:-ThF4-UFs mix-
tures appears suitable to permit use of a high
concentration of ThFs in melts whose freezing
point will prove acceptable for single-fluid
molten-salt breeder reactors. Oxide tolerances
of such mixtures are not entirely defined, but
laboratory experience and the MSRE operating
experience combine to suggest that no oxide
scavenger (such as ZrF,;) need be added. The
design basis fuel, accordingly, consists of 71.7
mole% "LiF, 16 mole% BeF:, 12 mole% ThF,4, and
0.3 mole% 2**UF.. The secondary coolant is to be
8 mole% NaF with 92 mole% NaBF4. Physical

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

FEBRUARY 1970

Grimes MSBR CHEMISTRY
properties of these fluids, estimated in most cases
from data on similar materials, seem adequate
for their use in MSBR.

Compatibility of the LiF-BeF:-ThFs-UF4 melt
with Hastelloy-N and with moderator graphite
seems assured, and operation of LiF-BeF2-ZrF,-
UF, melts in MSRE indicate very strongly that the
compatibility will not be adversely affected by the
consequences of radiation and fission of the ura-
nium, Compatibility of the secondary coolant, while
less satisfactory, appears to be adequate with
Hastelloy-N if salt purity is maintained.

Fission product Kr and Xe are virtually insolu-
ble in the fuel and can be removed, if the moder-
ator graphite is sufficiently impermeable, Dby
simple equilibration with an inert gas (helium).
Fission products with stable fluorides (Rb, Cs,
Sr, Ba, the lanthanides, and Zr) appear almost
entirely in the fuel as fluorides except as they are
lost through volatilization of precursors. More
noble fission products (Nb, Mo, Ru, and Te) are
virtually absent from the fuel, but plate (in ap-
preciable amounts) on the graphite and the Hastel-
loy-N, and appear to a surprisingly large extent in
the helium purge gas.

Chemical separations, of which reductive ex-
traction appears most attractive, for removing
uranium and protactinium from the fuel salt and
from each other have been demonstrated in small
scale laboratory equipment. Separations of lan-
thanides from the fuel are markedly more diffi-
cult, but reductive extraction of these elements
into molten bismuth appears possible.

While much research and development remains
to be accomplished there seems to be no funda-
mental chemical difficulty with design and opera-
tion of a single-fluid molten-salt breeder system.

ACKNOWLEDGMENTS

This research was sponsored by the U.S, Atomic
Energy Commission under contract with the Union
Carbide Corporation,

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NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

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