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OCR for page 37
Fluoride Salt Chemistry,
Partitioning, and System Corrosion
To provide a technical evaluation of the options identified for the
disposition of stored fluoride and flush salts, the chemistry of these salts
needs to be understood. The effects of radiolysis and the processing steps
required for remediation of these salts, as well as the corrosive
interactions of the salts with the piping, storage tanks, and other
components, also lead to important chemical changes in the salts.
CHEMISTRY RELEVANT TO THE PRESENT STATUS
As shown in Table 3.1, the composition of the flush salt in mole
percent is 65.9 lithium fluoride (LiF), 33.9 beryllium fluoride (BeF2), and
0.~8 zirconium fluoride (ZrF4~. The fuel salt has the mole percent
composition 64.5 LiF, 30.4 BeF2, 4.9 ZrF4, and 0.12 uranium
tetrafluoride (AFT. The LiF and BeF2 components were chosen to be
thermodynamically more stable than UF4, with a liquidus temperature for
the flush salt of about 460°C. The 4.9 mole percent ZrF4 in the fuel salt
reduces the liquidus temperature slightly but otherwise serves two
functions:
1. For the composition of the fuel salt, depending on the
temperature, zirconium may be reduced slightly more readily than
uranium (tending to minimize the formation of metallic uranium).
2. The ZrF4 component reacts more readily with oxygen than
UF4 and can serve as a getter to scavenge any oxygen impurity.
An unusual and favorable property of these salts is that their
solidification shrinkage for the given compositions is only about 2
percent.
37
OCR for page 38
38
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
TABLE 3. ~ Composition of the Fuel and Flush Salts
Fuel Salt Flush Salt
Current salt composition (mol YE
LiF 64.5 65.9
BeF2 30.4 33.9
ZrF4 4.9 0.18
Mass off ssile elements in salt big)
Uranium <33.2a 0.5
Plutonium 0.724 0.013
Current distribution of uranium and plutonium isotopes (wt tY)
232u 160 ppm
233u 83.92
234u 7.48
235u
236u
23su
239pu
24opu
Other Pu
2.56
0.104
5.94
90.1
9.52
0.35
75 ppm
39.4
3.6
17.4
0.245
39.4
94.7
4.8
0.50
aAssumes that at least 1.8 kg of uranium has migrated to the omegas system and at least 2.6 kg
of uranium is loaded onto the auxiliary charcoal bed.
SOURCE: Peretz (1996c, Table 1.3).
The fuel salt solids can be viewed as a matrix of closely packed
fluoride ions with the much smaller Be2+ and Li+ ions occupying
interstices. At their low concentrations, Zr4+ and the larger U4+ cation
contribute little to spatial requirements (Zachariasen, 1948~. Indeed, on
the basis of fluoride volume alone, the fuel salt density is estimated
(Robert Penneman, unpublished), using the data of Zachariasen (1948),
as 2.52 g/cm, close to the 2.48-g/cm3 density reported (Peretz, 1996c).
Chemical Consequences of Radiolysis
As a consequence of the radiolysis reaction, some cations of the
stored crystalline salt may be reduced to metal atoms. Peretz (1996c)
suggests that any of the metal atoms lithium, beryllium, uranium, or
OCR for page 39
CHEMISTRY, PARTITIONING, AND SYSTEM CORROSION
39
zirconium could be formed; this is uncertain because the reduction
potentials for the metals in fused salt at 500°C are not relevant in the
crystalline solid phases. Perhaps the identity of the metal atoms is not
important, since they can be considered simply as electronic point defects
in the crystalline salt phases, not metal particles.
Another important consequence of the radiolysis reaction,
besides the loss of gaseous fluorine and uranium hexafluoride (UFO) by
migration to the piping system, is that the highly reduced multiphase salt
has physical and chemical properties, such as an apparent reduction in the
solubility of various fluorides in the melt, that differ from the properties
of the original, single-phase fused salt. To better inform remediation
strategy, it is important to gain additional information on the (as yet)
poorly defined chemical properties of the remelted salts.
PARTITIONING OF URANIUM FROM THE SALT
As noted in Chapter 2, sampling of the gas phase near the end of
the off-gas vent line indicated high concentrations of F2 essentially
saturated with UFO vapor at the current temperature of the vent system
(~21°C). This gaseous product results from radiolysis reactions in the
solidified salt phase, and it would be prudent to assume that UFO has been
distributed to, and condensed in, all regions of the system, including the
freeboard volume in the upper regions of the drain tanks as well as areas
of restricted flow in the off-gas vent system preceding the auxiliary
charcoal bed (ACB). Due to the significant amount of alpha decay arising
from the 232U daughters that grow into the chain with a 1.9-year half-life,
radiolysis effects make it probable that a significant fraction of the initial
UFO may now be present as non-volatile lower uranium fluorides.
It is possible that back-migration of oxygen or water vapor from
the ACE into the vent line has occurred. This could result in the
precipitation of urany! difluoride (UO2F2) as one of the species currently
restricting flow in the vent line and in the production of hydrogen
fluoride (HF) gas. Today the ACB is isolated from the vent line.
The staff at Oak Ridge National Laboratory (ORNL) proposes to
clear the off-gas line by initially purging with an inert gas (helium or
OCR for page 40
40
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
argon), followed by evacuation to volatilize any UFO residing in the off-
gas system; this approach shouicI remove all UFO, but it would not
remove lower fluorides that may be present in the system.
Several chemical options can be considered for the removal of
nonvolatile uranium residues remaining after the initial pumping.
Considerable fluorination technology that has been developed over the
last 20 years at other Department of Energy (DOE) sites may be
applicable to the Molten Salt Reactor Experiment (MSRE) remediation
effort. These options may be of use both for clearing the piping and traps
in the off-gas lines and for recovering 233U from the remelted MSRE salt
fractions by volatility processing methods.
Concern exists about whether the solidified salt can be remelted
to form a homogeneous liquid since significant fluorine has been lost
from the original melt (Chapter 2~. To address this issue, Chapter 4
describes a recommendation to attempt to refluorinate the salt
components during a heating phase short of actual melting by addition of
HE gas in helium (He) at a slightly positive pressure. It is expected that
HF-He would oxidize all uranium or plutonium compounds to the IV
oxidation state at 400°C (just below the melting temperature) if adequate
permeability exists in the solidified salt. Additionally, refluorination can
be continued during progressive melting with a rock-melting, laser, or
cairod apparatus to ensure that significant quantities of fissile material do
not accumulate as a critical deposit. Hydrogen fluoride will not oxidize
uranium or plutonium to the hexafluorides UFO or PuF6, and molecular
fluorine at temperatures less than 300°C may not be effective in
producing UFO or PuF6.2
In order to generate UFO from UF4, especially at the temperature
existing in the off-gas piping, alternative fluorination agents should be
considered to boost the reaction rate. These would include atomic
As is well known to MSRE project personnel (Rushton et al., 1996a,b), any such
central heat source to induce localized melting need not be powerful enough to supply all
the requisite heat energy to the salts. The resistance heaters on the external tank wall could
be used to elevate the salt temperature.
2Fluorine reacts with UF4 to produce UFO in argon-neon mixtures at temperatures as
low as 12 K. Large-scale, efficient production of UFO may require higher temperatures or
ultraviolet irradiation (Margrave et al., 1976, 19773.
OCR for page 41
CHEMISTRY, PARTITIONING, AND SYSTEM CORROSION
41
fluorine, bromine pentafluoride (BrFs), chlorine trifluoride (ClF3),
fluorine dioxide (O2F2), dioxygen monofluoride (O2F), and krypton
difluoride (KrF2). All of these have significant advantages and
disadvantages, as discussed in Appendix B.
PLUTONIUM PARTITIONING BY FLUORINATION
Peretz (1996c, p. 3-24) referred to laboratory and pilot plant
studies involving fluorination of surrogate salt containing both uranium
and plutonium. The results showed that very little plutonium was
removed, even after most of the uranium was volatilized as UF6.3
Furthermore, the high-temperature fluorination yield of PuF6 was low,
and long fluorination times caused increased corrosion.
These results are consistent with the understanding that, because
PuF6 is a powerful fluoridating agent, little PuF6 would be expected to be
volatilized from the molten salt phase until all other volatile fluoride
gases (including all the UFO) are volatilized and driven out of the molten
salt solvent. Experience at Los Alamos (Mills, 1996) suggests that
volatilization of such small amounts of PuF6 (at the level of ~ 55 parts per
million [ppm] in the existing salt) is difficult, especially if plutonium
exists in the salt phase as a double salt with one of the major components
(e.g., LiF). Ambient-temperature fluorination using the reagent O2F2 is
also unlikely to be successful in removing plutonium from the bulk salt
(see Appendix B).
Based on these considerations, fluorination does not appear to be
usable for removing plutonium from the salt. The depleted salt will retain
the plutonium (and fission products).
3A 45-hour fluorination of 2000 liters of MSRE molten salt removed 216 kg of
uranium and left a uranium residue of 26 parts per million (ppm; Peretz, 1996c). No
uranium was volatilized during the first 7~/: hours while conversion of lower oxidation
state uranium to uranium pentafluoride (UFs) occurred. Subsequently, UFO was volatilized
at about 6-7 kg per hour. During the last 6 hours, fluorine utilization dropped to nearly
zero, and essentially no plutonium was removed (the final plutonium concentration was
1 10 ppm, compared to an initial value of 1 12 ppm).
OCR for page 42
42
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
NONFLUORINATION OPTION FOR PLUTONIUM
SEPARATION
If plutonium separation becomes essential, the following work at
Los Alamos might be the basis for development: Los Alamos
investigators (Ioel Williams, Los Alamos National Laboratory, personal
communication with Melvin Coops) have found that it is feasible to strip
small amounts of plutonium from molten salt electrorefining residues (a
eutectic salt of sodium chloride and potassium chloride [NaCI-KCI]) by
bubbling oxygen through the molten salt to precipitate plutonium (and
americium) as the insoluble dioxide. The salt is then evaporated for
recycle, leaving the oxide residues as a waste product. Since oxygen gas
has a low solubility in the molten salt, addition of sodium carbonate to
the molten salt has been found at Los Alamos to be a suitable way to
precipitate the plutonium-americium oxides.
If the MSRE molten salt is sufficiently fluid to be either filtered
or centrifuged to separate a precipitate of zirconium-plutonium oxide, the
method utilized at Los Alamos may be an applicable technique for
isolation of the small amount of plutonium present in the MSRE drain
salt. Addition of lithium carbonate (Li2CO3) to the salt after uranium
removal is complete may be a simple and effective method of adding
oxygen to precipitate plutonium oxide (Pu02), with zirconium oxide
(ZrO2) also likely.
SYSTEM CORROSION ISSUES
In considering the recommendation for in situ melting of the
stored salts (first the flush salt and then, if successful, the fuel salt) and
pressurization of the tanks to push the molten salt through the existing
(cleaned and heated) valves and pipes, it must be ascertained that these
Hastelloy N components have not suffered severe corrosion damage
during their use and storage periods. Hastelloy N was a good choice of
material initially, because the high-nickel alloy is least susceptible to
fluorination attack at high temperature (Lad, 1990~. Although fluorine
corrosion at high temperature is accelerated significantly by the presence
OCR for page 43
CHEMISTRY, PARTITIONING, AND SYSTEM CORROSION
43
of water vapor (HF is formed), in this case water vapor would also have
been removed from the gas phase by the formation of ZrO2 and UO2F2.
Any significant corrosion of the alloy at ambient (or warm, 200°C)
temperature would require the presence of an aqueous (or water vapor)
phase.
There is no evidence that system internals have been exposed to
moisture via any leak in the system. ORNI, corrosion researcher James
Keiser inspected one valve that had been plugged and cut out of the
system and found no evidence of observable corrosion attack. As one
possible means of observing on a gross scale the condition of the
Hastelloy and the salt inside the storage tanks, an optical fiber inspection
system could be considered. A cored salt sample taken from the center of
a storage tank could not be expected to contain the chromium, iron, and
other solutes that would indicate tank corrosion because any corrosion
products (particularly nonvolatile compounds or those that complex with
fluoride salts) would be localized to the immediate surface of the alloy.
Radiation-Induced Corrosion Questions
The 30-year-old drain tanks have been experiencing a radiation
field and radiation-produced compounds, and one can ask what effect
these compounds have on the integrity of the Hastelloy N walls. As noted
previously, there is about one atmosphere of excess pressure over the fuel
drain tanks. This is likely to be largely F2 combined with the saturation
pressure of UFO (and HF-O2 [molecular oxygen] if there has been any
leakage of moist air). Chapter 2 addresses the effects of alpha self-
radiation on decomposing, solid UFO. However, radiation effects on the
F2 gas could cause some reversal of UFO decomposition. Indeed, F2 plus
radiation, especially in the presence of 02, iS an aggressive reaction
mixture. Even at ambient temperature, at which F2 itself will not react,
enhanced chemical attack of the fuel tank wall might take place in a
radiation field, liberating molybdenum or chromium hexafluorides (MoF6
or CrF6) as volatile gases and possibly corroding the nickel by forming
Li2NiF6
The F2 and UFO gases are more reactive and corrosive than the
solid salt, and it is unknown at the present time whether these gaseous
OCR for page 44
44
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE;
species are in contact with the interior drain tank wall. Because the fuel
salt shrinks approximately 2 percent on solidifying, a gas space may exist
between the solid salt surface and the interior tank walls. Alternatively, a
coating or film of solid salt may be deposited, protecting the tank wall
from reactive gas attack.
Some laboratory-scale corrosion tests could be performed, but
these would take time and would be of only limited value, since it would
be difficult for such tests to duplicate stresses or other local conditions
such as pitting. The ideas4 presented below are offered for consideration
by ORAL staff to weigh the merits of obtaining this information against
the time and efforts required for meaningful results. Corrosion tests
would be conducted over irradiation times short compared with the 27-
year radiation exposure of the fuel tank walls, even though the integrated
dose could be similar. For example, samples of metal of composition
equivalent to the tank and thimble tubes could be prepared and sealed in
contact with a variety of gas environments, such as F2, F2+ 02, and 50-50
mixtures of UF4 and UFO. Then these samples could be irradiated with
gamma rays (for example, from the High Flux Isotope Reactor tHFTR]
spent fuel elements) or with a source of alpha particles to simulate the
history of the drain tank walls in the presence of radiation fields. Over
time, the samples could be tested for metal corrosion and for formation of
volatile UFO.
To provide perspective on the severity of these corrosion issues,
the fact that the tanks and associated piping continue to hold gases at a
pressure of more than one atmosphere after more than 25 years suggests
that there are no leaks in the metal confinement at this time.
4Many techniques, such as standard procedures in nondestructive testing, are not
practical due to the intensity of the radiation field at the tank wall (see Chapter 2),
necessitating the use of remote operations to access the drain tanks.
Representative terms from entire chapter:
fuel salt
