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OCR for page 50
Comments on Specific Separation
Technologies
Because separation of uranium from the fluoride salts is a key
step in the remediation plans, the pane} has considered in detail the
candidate technologies for this activity, and comments on them here.
FLUORINATION
The best candidate option based on current data and Molten Salt
Reactor Experiment (MSRE) status information seetus to be recovery of
the uranium fraction by fluoride volatility techniques. This technology is
thoroughly demonstrated and can be performed safely.
Although removal of uranium from the molten irradiated salt has
been done in the past, another question exists—can aged, solidified salt
be remelted and brought to a homogeneous melt condition? If the answer
is yes, then the chemistry and process of fluorination can proceed given
adequate vessel integrity. This also assumes that the piping is unblocked
and free of leaks. Lower-temperature fluorinating agents may be backups
if vessel integrity indicates that high-temperature direct fluorination with
fluorine may be dangerous (see Chapter 3 and Appendix B). Salt
hydrofluorination, discussed below, may be advantageous prior to
melting.
Small-scale laboratory tests in gas-tight enclosures with samples
of actual salt, even if not truly representative, are important to evaluate
whether the salt can be remelted and, if so, under what conditions. As
noted elsewhere (Appendix C), collecting samples poses a possible radon
daughter contamination hazard. Prior to obtaining an actual salt sample,
tests with irradiated surrogate salts should be continued.
50
OCR for page 51
COMMENTS ON SPECIFIC SEPARATION TECHNOLOGIES
51
The possibilities of direct fluorination, hydrofluorination, and use
of alternative fluorinating agents are discussed in further detail below.
Direct Fluorination
One technical approach (Peretz, 1996c, 3-7,8), as elaborated by
Rushton et al. (1996a), involves initially forming a small puddle of liquid
salt by partially melting a small area on the surface of the fuel salt and
treating the slurry (melt and precipitate) with fluorine gas (F2), a mixture
of hydrogen fluoride and hydrogen (HF-H2), or a mixture of hydrogen
fluoride and helium (HF-He). if this produces a clear melt, then melting
would be increased slowly. Rather than first solubilizing all the uranium,
use of fluorine results in removal of uranium hexafluoride (UFO) as the
melting proceeds.
Even though hydrogen would be produced in the restoration of a
metal atom site into a cation, there is concern that introducing an HF-H2
mixture would be hazardous because of the fluorine sites existing within
the salt. Diluting the HE in helium could lessen the hazard associated
with hydrogen gas, which is introduced to control corrosion. It is possible
that HE treatment would remove the reducing sites by reacting with the
metal atoms. Possible uranium trifluoride (UF3) precipitation is unlikely
because of its solubility in the salt, although local solubility excess could
occur.
Use of an internal heat source for remelting the solidified MSRE
salts directly within the two drain tanks has been suggested by Oak Ridge
National Laboratory (ORNL) staff. This heat source would be used in
parallel with the externally mounted tank heaters to ensure that melting
initiates in the center of the tank and progresses outward. It is believed
that this approach would permit a single puddle to be formed in the
center of the tank. This would have two benefits: the melt could be
monitored for insoluble species that might be encountered on remelt, and
a solid salt coating would be maintained against the walls, with the salt
serving as its own container as long as possible.
This puddle would be fed a fluorinating gas stream continuously
to reoxidize any uranium in the salt to uranium tetrafluoride (UF4) and
UFO, as well as to stir the melt. Because of the hot fluorine and HE that
The hydrogen is introduced (Peretz, 1996c) to adjust the redox potential.
OCR for page 52
52
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
are expected to be used for this purpose, elements that could form
volatile fluoride gases (i.e., the hexafluorides of chromium, molybdenum,
tungsten, tantalum, and niobium tCrF6, MoF6, WF6, TaF6, NbF6] and
carbon fluorides) should not be used in the central heater assembly. For
this purpose, probably the best material is pure nickel or Hastelloy N.
Electrical heating is assumed, and a good thermal conductor will be
needed to fill the gap (between the resistance element and the tubular
nickel jacket) that does not short out the resistance unit. Several suitable
choices of material are available.
Hydrofluorination
Another approach, which might precede the one just described'
could involve the following steps: after testing the tank with vacuum and
helium purge to remove any volatile uranium fluoride and fluorine, some
modest pressure of HE or F2 could be applied to the warmed (but not
melted) salt to permit diffusion to decrease the residual reducing
equivalents in the fuel salt. Considerable time is available even for slow
reactions. Because HE does not oxidize uranium beyond the TV oxidation
state, this hydrofluorination procedure would be primarily a measure to
ensure that any uranium in diffusive contact with gas in the system is not
reduced below that state.
Alternative Fluorinating Agents
As discussed in Chapter 2, radiolysis of transported UFO would
produce a variety of lower uranium fluorides' some of which can be
refluorinated easily to UFO and some that might not react at temperatures
below 500°C unless fluoridating agents stronger than F2 are used (e.g.,
bromine pentafluoride [BrF5] and chlorine trifluoride [CIF3], as discussed
in Appendix B). This is a critical point, since much of the condensed UFO
and associated degradation products probably are present in the piping
and upper regions of the drain tanks (see comments 1 and 2 in Chapter 4~.
Lower-temperature fluorination might also be important in evaluating
vessel integrity for the fluorination step. Removal of Missile material from
the MSRE system will require a stronger fluorination treatment than
melting with simultaneous hydrofluorination by use of HF-H2 mixtures.
Several refluorination steps may be needed, such as (~) partial melting
OCR for page 53
COMMENTS ON SPECIFIC SEPARATION TECHNOLOGIES
53
with simultaneous HF-argon or HF-He treatment to convert the lower
fluorides or metal to UF4, followed by (2) treatment with F2 or BrF5 in
argon or helium at melt temperature to convert UF4 to volatile UFO
(which would then be trapped in sodium fluoride traps); followed by (3)
treatment with BrFs or KrF2 (krypton difluoride) to remove deposits from
the cooler surfaces, valves, and unheated plumbing constituting the
balance of the system.
Monitoring the effectiveness of the proposed refluorination
treatments may be possible without actually penetrating the containment
vessels, by inspecting the vessel walls with collimated precision gamma
spectroscopy. Core samples of the solid melt may provide valuable
information on both the amount of segregation that has occurred and the
state of reduced material present in the tanks, if the radon daughter
contamination (Appendix C) hazard can be managed.
ELECTROREFINING
Preliminary information was provided to the pane} concerning the
possible use of the Argonne National Laboratory (ANL) electro-
metallurgical process for treatment of MSRE salts. The panel believes that
its application to MSRE salts should not be considered unless fluorination
treatment fails and only after use of an electrorefining system has been
demonstrated on MSRE salts. Pilot plant demonstration of this technique
ton chIoride-based pyrochemical processing of spent fuels Tom the
Experimental Breeder Reactor (EBR-II) is under way. The National
Research Council (NRC, 1995) Committee on Electrometallurgical
Treatment of DOE Spent Fuels has recommended that additional possible
applications of this technique be deferred until the demonstration on
EBR-~l spent fuel has been shown to be successful. Consistent with this
recommendation, electrometallurgical treatment of MSRE salts should be
viewed as a future alternative pending further testing and development.
The current ANL demonstration is being performed by use of
chloride salt electrorefining (ER). The proposed treatment of MSRE salt
would use fluoride salt ER, with a liquid bismuth cathode to isolate the
uranium and plutonium from the lithium fluoride-beryIlium fluoride-
zirconium fluoride (LiF-BeF2-ZrF4) salt solvent. Although the equipment
is similar, the chemistry is not. Since the kinetics, thermodynamics, and
OCR for page 54
54
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
solubilities of the fluorides are not the same as those of the chlorides, the
proposed fluoride process has to be demonstrated satisfactorily in the
laboratory before any commitment to the technology can be considered.
Insufficient process data are available to make an evaluation of this
concept at present.
The ER process necessarily would be carried out in an inert
atmosphere by remote operation behind heavy shielding. No suitable
facility of this nature now exists at ORNL. It is too early to estimate what
the costs of such an operation might be, but facility costs on the order of
tens of millions of dollars are not unreasonable. Use of the ANL facility
in Idaho faces major problems of shipping and of coordinating operations
at two separate sites in addition to problems of the technology.
For these reasons, the pane} considers ER a possible backup
technology, but it is too early to give any further consideration to its
application for MSRE cleanup.
DISTILLATION OF MOLTEN SALT
Very little information was provided on distillation of the molten
salt. The few data concerning this alternative that are discernible from
basic chemistry imply that this operation would not provide clean
separations because of the wide range of vapor pressures for the salt
constituents. It would also require very high temperatures, up to
approximately 1500°C; thus, serious containment problems would arise.
It would appear that both process (or flow sheet) conditions and process
equipment would have to be developed, and the potential benefits are not
apparent. The panel recommends that consideration be given to dropping
this process from the list of alternatives unless significant and favorable
additional information becomes available.
AQUEOUS DISSOLUTION AND SEPARATION
Another alternative is to dissolve the 4650 kg of fluoride salt in
an aqueous solvent and isolate the uranium and plutonium by known
aqueous processing techniques. Because the separation of uranium and
plutonium from fission products and other ions is well demonstrated for
OCR for page 55
COMMENTS ONSPECIFIC SEPARATION TECHNOLOGIES
55
precipitation, solvent extraction, and ion exchange, there is little danger
that such separations cannot be accomplished. Aqueous separation is also
the only ensured way to isolate the plutonium if that is an objective.
However, getting the salt into a soluble form suitable for processing
without segregation may be a major challenge. It may also be a
significant challenge to dissolve the plutonium if any of it has aged to the
reduced state. Some plutonium compounds become refractory upon
aging, and although soluble when first formed, they become insoluble
with aging.
If the aqueous work is done off-site, the salt removal (without
segregation) and shipping concerns must be addressed as well as the
technology for the dissolution and feed preparation steps. It would also
be necessary to resolve the question of disposal of the fluoride-bearing
liquid waste.
If aqueous processing is done on-site, questions of what to do
with the waste still remain, except that the volume of waste generated
would be much greater than from nonaqueous means. This larger volume
is derived from the low solubility of MSRE fluoride salts in dilute
mineral acids, which implies that significant clilution with water is
required. The neutron-moderating capability of water poses a criticality
concern, discussed in more detail below. Facilities at ORNE that were
used for aqueous processing of 233U in the past may not be suitable for
the current mixture of U. 32U, and plutonium. New facilities would be
very expensive.
Criticality Concerns in Aqueous Processing
The issue of criticality prevention would have to be addressed in
dissolving the salts. If it is possible mechanically to retrieve the salt from
the existing MSRE system for external chemical treatment, dissolution in
a favorable geometry for aqueous treatment would be required to ensure
2A favorable geometry is said to exist when a container or piece of equipment cannot
hold enough fissile material to be critical regardless of enrichment, concentration, or
amount of water-equivalent external reflection. For example, a long thin cylinder 11.7 cm
in diameter and of unlimited length (an "infinite cylinder") will be subcritical if filled with
233U in nitrate solution of any concentration (i.e., with any amount of uranium that can
actually be dissolved in an aqueous or organic fluid) with a water reflector. The term safe
geometry, although often used synonymously, is discouraged because such a container
OCR for page 56
56
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
criticality safety. This problem arises from the fact that BeF2 present in
the salt is readily soluble in many aqueous solvents, whereas LiF and
ZrF4 (zirconium fluoride) are not. In any mineral acid (for example, 3
molar nitric acid) the fluoride ion produced by BeF2 dissolution would
cause immediate precipitation of all uranium, plutonium, and lanthanide
elements as insoluble fluoride salts. These precipitates have high density,
are insoluble in nearly all mineral acids, and could aggregate rapidly to
form a critical mass. As little as approximately 600 g of 2 U in aqueous
solution could go critical if the dissolver vessel is not configured as a
favorable geometry (APSE. ~ Working Group, 1993~.
Fluoride Removal
The amount of fluoride present in MSRE fluoride salts would
cause major complications in current aqueous processing methods, such
as a solvent extraction process (e.g., the plutonium and uranium recovery
by extraction, or PUREX, process) or ion-exchange systems. Uranium
fluoride can be dissolved by metathesis of the fluoride to the hydroxide
followed by acid dissolution, or by reacting UF3 or UF4 with warm borate
salts to volatilize fluorides as boron trifluoride (BF3) gas. Another
possible approach is addition of aluminum nitrate to sequester the
fluoride and provide the concentrated aqueous salt concentrations
necessary to put uranium and plutonium into the solvent phase. In either
case, the fluoride ions present in the solution must be removed prior to
continued aqueous chemical processing, whether by ion exchange,
solvent extraction, or precipitation.
Conclusions on Aqueous Processing
Because aqueous processing is the basic technology for
separations work in this field, it was reviewed as a candidate for
application to MSRE salts. Because of the extensive experience
(worldwide) with reprocessing nuclear fuels of many materials and
compositions, the pane! believes that a workable flow sheet could be
developed, but due to complications introduced by the large amount of
would not necessarily be "safe" (i.e., subcritical) if other nearby containers (of fissile
species) were also present.
OCR for page 57
COMMENTS ON SPECIFIC SEPARA TION TECHNOLOGIES
57
fluorides in the system, the result would be very complex, would produce
great amounts of waste, and would necessitate new facilities because of
the corrosive nature of the solutions required. Such chemical processing
would have to be performed not only in critically safe vessels, but also by
remote means behind heavy shielding, because of the fission products
present. For these reasons, the pane} considers aqueous dissolution to be
one of the costliest and least desirable process options currently under
consideration.
STABILIZATION TECHNOLOGIES
Most of the stabilization technologies listed in Peretz (1996c)
involve shipping material off-site, making both transportation and
acceptance issues of concern. Because it might not be permissible to ship
the salts in their present unstable form, stabilization should be planned as
a near-term necessity. Stabilization (as with a chemical getter) may also
be required for on-site interim storage and may be possible in association
with another operation.
Some nuclear criticality safety issues are specific to a particular
processing route, such as aqueous processing. As another example,
solubilizing the entire salt mass prior to any uranium removal in a
fluorination process would pose a greater hazard than remelting and
fluoridating a smaller mass. Further remarks on nuclear criticality safety
are offered in the next chapter.
Representative terms from entire chapter:
msre salts
