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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
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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.
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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
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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
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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
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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
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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.
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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: