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Nuclear Wastes: Technologies for Separations and Transmutation APPENDIX D SEPARATIONS TECHNOLOGY-ADDITIONAL INFORMATION INTRODUCTION All current processing plants for reactor fuel have similar steps. The first is the separation of the fuel from a protective sheath called cladding. This sheath is metal; Zircaloy, aluminum, and stainless steel are the usual materials. These materials are alloys, and the nature of the minor elements must be considered. Zircaloy cladding may be made with a few percent of tin or niobium. The aluminum cladding may contain silicon, and stainless steel cladding is usually made of iron with 18% chromium and 8% nickel. Fast reactor cladding is essentially iron and chromium with very little nickel present. In processing of commercial spent oxide fuels, Zircaloy is removed by mechanical chopping of the fuel rods into segments, followed by dissolution of the spent nuclear fuel in nitric acid. Zircaloy may also be dissolved away from metal fuel with nitric and hydrofluoric acids. In this zirconium cladding dissolution procedure, the dissolver solution is later reacted with aluminum nitrate to form fluoride ion complexes in order to prevent excessive corrosion of downstream process equipment and to supply a salting agent to promote the extraction process. This process generates large volumes of waste per unit volume of fuel processed. The mechanically removed hulls may retain some activity but are a minor process and waste volume problem. Aluminum cladding is often dissolved in nitric acid catalyzed with mercury; alternatively it may be dissolved in sodium hydroxide sodium nitrate solution. The caustic solution is then filtered to recover the actinide fraction, or made acidic to produce the feed solution. In some processes like REDOX, the aluminum nitrate serves as a salting agent for the extraction of plutonium and uranium. Unfortunately, this aluminum nitrate constitutes a waste stream that is ten times larger than the fission products themselves. The usual approach with stainless steel is to shear the cladding and dissolve the fuel oxide in HNO3, leaving the steel hulls in solid form for final disposal. The stainless steel is not dissolved because the elements in the alloy interfere with the aqueous separation processes, but it can be dissolved with anodic corrosion methods when using pyrochemical separation techniques. PUREX feed is made by leaching the fuel from the cladding, thus leaving nitric acid as the only salting agent and the actual fuel as the solute. Nitric acid is easily recycled for reuse with simple distillation and concomitant minimization of wastes. The process may be used with added salts in the feed if desired, but the optimum waste stream is not produced this way. In the case of the pyroprocesses conducted electrochemically in molten salt, the metallic sheath (usually stainless steel) might be removed from the fuel by anodic dissolution. The fuel oxide dissolves in the molten salt. If the salt is lithium chloride the oxides may be converted to the
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Nuclear Wastes: Technologies for Separations and Transmutation metals by reaction with lithium metal and prepared for further electrochemical processing. The oxygen present reacts to form lithium oxide, which may be removed from the salt by electrolysis with a carbon anode to regenerate the lithium metal at the cathode for reuse, and carbon dioxide which is discarded as waste. This avoids the formation of excessive solid process wastes. Molten salts and alloys have a long history of use in the processing of nuclear weapons materials. They are used in processes for production of lithium metal and lithium compounds as well as of uranium and plutonium metals. Essentially all processes for the manufacture of fluorine and intermediate uranium compounds for uranium hexafluoride production are pyrochemical. Many of the recycle processes for uranium and plutonium, including residue recovery, are pyrochemical. There are no aqueous chemical analogues for many of the steps done by pyrochemical means, but there are advanced pyrochemical techniques that can replace many traditional aqueous processing methods. The single-stage separation factors in pyrochemistry can be large for equilibria between liquid metal and molten salt phases, and cascades are not usually required for fuel recycle. Multistage equipment using short-stage-time centrifugal contactors originally designed for aqueous systems is being developed for those molten-salt/molten-alloy processes that require a high degree of purification. High-temperature liquid-liquid extraction equipment is similar in concept to that used for aqueous systems. There is a large body of experience with the use of aqueous solutions and organic extraction phases for large separation cascades. There has been success using large and small separation factors, down to a stage separation factor of 1.002. This experience should be applicable to many of the proposed high-temperature molten-salt and alloy systems that generally show good separation factors for isolation of actinide elements from fission product residues. AQUEOUS PROCESSES Some insoluble products result from high burn-up fuel dissolution in nitric acid. The solids may be complex fission-product compounds or insoluble reaction products. High-fired pure plutonium oxide does not dissolve completely. Such solids would typically be removed from the solutions by filtration or centrifugation for further treatment or discard to waste. Both of these solid–liquid separation methods can be designed for easy recycle of valuable materials. The radioactive rare gases krypton and xenon can be removed in the off-gas system, perhaps by cryogenic absorption methods. Iodine can be chemically trapped from the gas phase during dissolution. A process to separate volatile ruthenium or technetium oxides could be included at the time of initial dissolution of the spent fuel. After dissolving the fuel, the next step is to separate the uranium and plutonium from the very radioactive fission products and the higher actinides. With the uranium in the uranyl form (+6) and the plutonium in the +4 oxidation state, these elements may be separated from all other materials of concern to almost any degree desired using a multistage cascade that yields very low losses of uranium and plutonium to waste. PUREX accomplishes this task very well, but in most plants the actual recovery of uranium and plutonium is not as good as theoretical design
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Nuclear Wastes: Technologies for Separations and Transmutation indicates. This is frequently due to the formulation of polymeric species during the extraction process. Solutions to this problem are now available and could be incorporated into the process. The Butex process, based on the extraction solvent dibutoxy diethylene glycol (dibutyl carbitol) has been used in the U.K. weapons complex for the purification and isolation of plutonium. This flammable ether is used much like the ''hexone" of the oxidation-reduction chemistry process (REDOX). A series cascade of a PUREX stage and a Butex stage has been used to recover and purify uranium to the part per million (ppm) level for four decades at the Y-12 Plant at Oak Ridge. Tributyl phosphate (TBP) is used as the PUREX extractant.1 It suffers from hydrolysis and radiolysis products that complicate the product recovery step that is carried out by stripping plutonium and uranium from the organic phase with dilute nitric acid. The solvent must be continuously purified to achieve the high recovery and decontamination levels needed. TBP is purified in PUREX plants by removing deleterious degradation products that result from both radiolytic and chemical processes; contact with sodium carbonate solution is used. Removing the degradation products from the TBP generates large volumes of contaminated wastes that must be eventually treated for disposal. The equipment used for radiochemical separation processes needs to be designed for extreme reliability and with very low maintenance. This means that much of the chemical equipment has to be specifically designed for the operation to be performed, and generally, commercial items are not satisfactory. Much of the process equipment needed has no commercial use outside the nuclear industry. The extraction equipment used for liquid–liquid separation processes is however, to a first approximation, independent of the solvents used. Thus, new processes that may be developed using, for example, liquid ion exchangers, phosphine oxides, phosphoric acid derivatives, or solvent systems based on amides, may entail only simple changes in operating conditions rather than new plant equipment. In aqueous plants, extraction and stripping can be conducted in several different types of equipment. Most plants use either centrifugal contactors, mixer-settlers, or pulsed columns, occasionally some of each. Centrifugal contactors are gaining favor, because they are the most compact and fastest types of solvent extraction systems and thus minimize shielding costs, increase plant throughput, and reduce radiation damage to process reagents. Some plants employ different types of equipment in different stages of the process. To minimize radiolytic damage to solvents, equipment with short contact times (such as centrifugal contactors) may be selected for first-cycle extraction, where the greatest fission product, radioactivity, is present. Simpler equipment, such as pulsed columns or mixer-settlers, is satisfactory for second and third cycles, where radioactivity is several orders of magnitude lower. In Europe, where there is considerable industrial experience in aqueous processes (Butex process is no longer used), many improvements have already been realized, e.g., Centrifugation of fines installed. 1 The reactions for TBP are: (a) UO2(NO3)2(aq) + 2TBP(o) = UO2(NO 3)2(TBP)2(o); (b) Pu(NO3)4(aq) + 2TBP(o) = Pu(NO3)4TBP2(o)
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Nuclear Wastes: Technologies for Separations and Transmutation Plutonium recovery in UP3 99.7%. TBP withstanding fuel with burn up of 4GWDT especially when using pulsed columns. Centrifugal contactors have already been used in second cycle operations. Bismuth Phosphate Process In 1941 it was known that plutonium has multiple oxidation states. A greatly expanded investigation of the aqueous chemistry of plutonium on the tracer level was initiated, which included separations methods based on precipitation, organic solvent extraction, and other approaches. It was found that plutonium coprecipitates in its reduced states with lanthanum fluoride, but not in its oxidized states. It coprecipitates with iron hydroxide and zirconium phosphate, which are typically gelatinous, hard-to-filter precipitates. A precipitation method in which most other elements would not follow plutonium was developed based on the ability of zirconium phosphate precipitate to carry plutonium in the +4 oxidation state but not in the +6 state. The reducing agent was NaNO2, and the oxidizing agent was NaBiO3. It was found that the addition of sulfate to uranium solutions formed complexes with the uranyl ion and suppressed precipitation of uranium phosphate, while NaNO2 did not reduce UO2+2 to U+4. NaBiO3 oxidized Pu+4 to PuO2+2, and the formation of plutonyl ion prevented precipitation of plutonium phosphate from oxidized solutions. The process was soon converted to BiPO4 precipitation, since it performed as well as Zr 3(PO4)4 as a carrier, and the precipitate was granular, greatly simplifying isolation. The NaBiO3 oxidation step provided an internal supply of Bi+3 for scavenging the unwanted fission products, which were precipitated from the oxidized solutions as phosphates prior to carrying Pu+4 on bismuth phosphate. This was the first large-scale radiochemical separation to exploit oxidation-reduction processes for both the purification and concentration of a single element from many other contaminants at these low concentrations. It is notable that the Bismuth Phosphate Process had a scale-up factor of 108 from lab to plant and produced 99.9% pure plutonium product at a 97% chemical yield, with a decontamination factor of better than 107 from the radioactive fission products (Lawroski, 1955). The Bismuth Phosphate Process demonstrated the practicality of obtaining decontamination factors in the 107 range, which was needed for manual handling of the plutonium product in later fabrication steps. The only purity difference from current processes was in the chemical yield, which was slightly less. REDOX Process The Bismuth Phosphate Process recovered only plutonium from the irradiated uranium feed. The REDOX process, based on hexone extraction, was developed to simplify the remote mechanical operations. This method was called the "REDOX process" because of its use of oxidation-reduction separations chemistry. It exploited the ability of methyl isobutyl ketone (hexone) to extract both plutonium and uranium (and also neptunium) from oxidizing solutions
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Nuclear Wastes: Technologies for Separations and Transmutation (Hill and Cooper, 1958). This counterflow extraction process was carried out in stainless steel equipment and used column cascade extraction technology. The aqueous feed consisted primarily of molar nitric acid that contained the fission products and transuranic (TRU) elements as nitrate salts. Sodium dichromate oxidant was used to ensure that both the plutonium and uranium were in the hexavalent state (as uranyl and plutonyl ions). In a plant with enough stages, methylisobutylketone (known as hexone or MiBk) extracts these ions from the fission products virtually completely. In this process essentially all of the radioactive fission products and the excess oxidant and its reduction product were rejected to waste in the first process step, minimizing oxidation and radiolysis problems of the solvent and radiolysis of downstream reagents needed for plutonium isolation (Culler, 1956:201-211). Recovery of plutonium from the organic solvent was achieved by contacting the organic stream with a dilute nitric acid aqueous phase that was heavily salted with aluminum nitrate and contained a moderately strong reducing agent (Culler, 1956:172-194). Ferrous ion was selected to reduce PuO2+2 to Pu+3, but not UO2+2, which remained in the organic phase. The uranyl ion was stripped from the solvent with dilute nitric acid later in the process. Like the Bismuth Phosphate Process that exploited the differences in solubility of plutonium phosphate between Pu+4 and PuO2+2 oxidation states, the REDOX process utilized changes in the plutonium oxidation state to effect a separation from fission products and other actinides. PUREX Process Following the Manhattan project, many solvent systems were investigated for use in plutonium extraction (Culler, 1956:172-194). Much effort was concentrated on the alkyl phosphates after it was found that TBP had useful extraction characteristics for both plutonium and uranium recovery. None of the other candidate solvents at hand at that time had characteristics significantly better than TBP for plutonium and uranium extraction. An advantage of TBP was that it allowed use of 8 M nitric acid as the process salting agent, whereas most other extractant systems required aluminum, magnesium, or some other highly soluble nitrate in concentrated solution. TBP was also available in large quantities in a very pure state at a reasonable cost. When diluted to 30% by volume with kerosene, TBP extracts Pu+4 and UO2+2 with distribution coefficients (Kd) of about 20 from 6 to 8 M nitric acid. It is not as flammable as hexone and is highly immiscible with nitric acid solutions. It is reasonably stable to radiolysis and hydrolysis during processing and can be purified from decomposition products by contacting it with an aqueous sodium carbonate solution. Because of its chemical process simplicity, TBP became the successor extractant to hexone for aqueous separation processes. The PUREX process resembles the REDOX process in that essentially all fission products are rejected at the first stage of the extraction sequence. The feed is a much simpler solution of uranyl nitrate and the other elements in 6 to 8 M nitric acid, with a trace of nitrite present to stabilize Pu+4. Only uranium (as UO2+2), plutonium (as Pu+4), and neptunium (as NpO2+2) are extracted from the nitrate feed; other salting or special oxidizing agents are not used. The uranium and plutonium are decontaminated together, and the partitioning of
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Nuclear Wastes: Technologies for Separations and Transmutation plutonium from uranium is done by hydroxylamine reduction of Pu+4 to Pu+3, which is then back-extracted into 6 M nitric acid while UO2+2 remains in the solvent phase. The decontaminated Pu+3 is removed from the organic stream, free of uranium, and the uranium fraction is recovered by back-extraction in very dilute nitric acid. Because PUREX does not need process salting to work efficiently and both solvent and aqueous reagent streams can be reclaimed for recycle, the volumes of waste fluids are greatly reduced. The fission product fractions can be discharged in concentrated form to holding tanks for radioactive decay prior to final vitrification and disposal. The process system works very reliably and produces a plutonium product with a decontamination factor from fission products of greater than 107. The uranium fraction has a similar decontamination factor. NONAQUEOUS PROCESSES Fluoride Volatility Processing Uranium hexafluoride production processes were initially developed to produce feed for the gaseous diffusion process for uranium enrichment and became routine on the multiton per day scale. The final purification of the uranium from ore is usually by solvent extraction, and the recovered uranium oxide is reduced to the dioxide, hydrofluorinated with hot anhydrous hydrofluoride, and then fluorinated with fluorine from a nonaqueous electrolytic cell. When it was found that plutonium also formed a volatile hexafluoride with characteristics very similar to uranium hexafluoride, it became apparent that plutonium and uranium might easily be recovered from irradiated fuels by volatility processing. Few other elements form volatile fluorides. Exceptions among the fission products are molybdenum, technetium, ruthenium, and tellurium. Tellurium (as Te2F10) is the only long-lived fission product closely following UF6 chemistry, while iodine (as IF5) follows UF6 chemistry to a lesser extent. However, both elements can be oxidized to higher oxidation states (TeF6 and IF7) that are much more volatile than UF6 and PuF6, and they can be separated from the uranium product by simple distillation methods. A process was demonstrated for purifying plutonium residues by fluorinating PuO2 in a fluidized bed reactor with fluorine at 400° C to form PuF 6. After purification the PuF6 was decomposed to PuF4 and F2 in a thermal decomposition column (Hyman et al., 1956). This system worked satisfactorily but was not put into production. More recently, a development effort was conducted through pilot-plant testing to demonstrate the use of fluorine oxide, O2F2, to convert PuO2 to PuF6 at ambient temperatures in simple reactor systems. Isolation of plutonium or uranium from bulk impurities or fission products by volatility methods has been demonstrated to be a practical approach that could be scaled to industrial levels. Fluoride volatility processing can also be used to recover neptunium (as NpF6), but there are no known hexafluoride compounds of americium, curium, or any of the other actinide elements heavier than plutonium. Therefore, its proposed use as a recovery technique for actinide elements from molten fluoride salt systems appears limited to volatilization of uranium,
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Nuclear Wastes: Technologies for Separations and Transmutation neptunium, and plutonium among the actinides. An alternative isolation method would be needed for the other actinide elements. Uranium and plutonium hexafluorides are powerful fluorinating agents that require use of nonreactive materials of construction. Monel, Inconel, and certain aluminum and copper alloys have been used successfully over the years in UF6 process systems, but shaft seals and packing glands pose significant challenges. Fortunately, proposed process systems have few moving parts. PuF6 is susceptible to internal radiolysis, with a significant decomposition rate to PuF4 and F2. The reaction is reversible, and in a closed system there would be an equilibrium composition dependent on pressure and temperature. Therefore, means to refluorinate the PuF4 residue must be built into any system handling PuF6 in quantity. Neutron emission occurs from PuF4 and PuF6 by (α, n) reactions on fluorine, and neutron shielding for personnel must be included in the design of all plutonium fluoride systems. Large-scale fluoride volatility plants for spent nuclear fuel reprocessing are certainly feasible, but none have been built in Western countries. A pilot plant utilizing fluoride volatility as a component of the overall plant system has reportedly been constructed in Dimitrograd, Russia, for the processing of fast-reactor demonstration fuels. Little information has been released on the utility of this prototype facility or on the economics of volatility processing under the conditions imposed by spent fuel. The existing large-scale fluorination systems that are used in the preparation of pure UF6 for the separation of the isotopes of uranium obtain very good separation from most elements. There have been difficulties with the separation of the very small amounts of molybdenum, technetium, neptunium, and plutonium that have been present in the feed derived from the recycle of reactor fuels. Technetium progresses very slowly through the diffusion cascades and is a constant problem once the system is contaminated with it. Tungsten and molybdenum hexafluoride move readily in the diffusion plant and can slightly contaminate the product. Molten-Salt Processes Two basic types of molten-salt systems have been used in separations: (1) the mixed fluoride salt fluid that was used as a coolant and a homogeneous fuel and blanket system in the Molten Salt Reactor Experiment, and (2) the simple chloride eutectic salts that are used as an ionic solvent in pyrochemical spent-fuel reprocessing systems that are intended for use with highly irradiated metallic reactor fuels. There are preliminary results on the reduction of oxide fuels with molten lithium and lithium chloride. Between 400 to 600° C temperatures, lithium metal in conjunction with lithium chloride reduces actinide oxides completely. The resulting actinide metals can then be processed as if they were irradiated metallic reactor fuels. In the Molten Salt Reactor Experiment application, the molten fluoride salt consisting of BeF2, 7LiF, ThF4, and UF4 was used as the working fluid. Molten salt provided a fertile thorium blanket, a neutron multiplier, the fissile fuel, and the reactor coolant, as well as much of the reprocessing solvent. The molten salt served as a suitable solvent for the bred 233Pa, which was extracted from the salt phase into liquid bismuth by reduction with a controlled
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Nuclear Wastes: Technologies for Separations and Transmutation concentration of lithium metal. The fission products were removed from the flowing salt loop by processing a small stream diverted from the main coolant loop with a higher concentration of 7Li metal. The purified salt was returned to the coolant stream after processing. The uranium was isolated as the volatile hexafluoride after fluorination with hydrofluoride and fluorine. The iodine was removed as hydro-iodine gas. The fate of the noble metals in the reactor was not resolved; they probably remained in metallic form as a suspension in the salt. One of the desirable features of using ionic molten inorganic salts and molten metals for fuel processing is that both phases are inherently resistant to radiation damage effects. The main advantage of pyrochemical processing is the ability to chemically separate "fresh" reactor fuels that have very high concentrations of fission products, with their associated high decay heat output and intense radiation emission. In fact, the decay heat may offer a significant advantage, since pyrochemical processing typically is performed at process temperatures of 500 to 800° C. Current pyroprocesses typically utilize the chemical-free energy between the molten element and its lowest oxidation state as an ion in a molten-salt solvent in a molten-chloride salt phase. Pyrochemical extraction chemistry does not have the complications of hydrolysis and chemical instability of aqueous extraction chemistry. Much of the needed thermodynamic data is at hand or is easily derived. Key to the evolving technology is the use of lithium chloride as part of the salt mixtures and lithium metal as a reducing agent. Magnesium chloride, copper chloride, and cadmium chloride may be used as selective oxidizing agents. The salt transport process (Steunenberg et al., 1969), a pyrochemical method for recovering actinide elements from spent fast-reactor fuels, is an example of a sophisticated pyrochemical system. In this process, plutonium and uranium are recovered from fission product residues and other spent reagents. The basis of the separation is the transport of actinide metals between two molten alloys of magnesium, one containing copper and the other containing zinc. The alloys are chemically connected by means of an ionic molten-salt solvent, which permits movement of ions between the alloys. The driving force of the process is the difference in the thermodynamic activity of plutonium and uranium in the two alloys. At 800° C equilibrium is established rapidly, and in only a few stages plutonium can be separated from uranium with a decontamination factor of about 100, while the fission products are readily isolated from the actinide elements present. The noble element fission products are retained in the copper–magnesium donor alloy, and the reactive fission products (cesium, strontium and the rare earth isotopes) remain as ionic species in the transport salt phase. No external electrical potential is needed in this process. The separation is accomplished by oxidation of the plutonium by the chloride solvent salt at the copper–magnesium interface and reduction of the plutonium chloride by the magnesium alloy at the magnesium-zinc interface. Mg+2 and Pu+3 move in opposite directions through the salt bridge. Processes of this type can be readily designed to effect separations between chemical families of elements and can also provide separations between individual members of chemically similar families of elements. This high-temperature REDOX separation technique has yet to be developed to its full potential. It appears to be well matched to reprocessing of metal fuels used in fast reactors, where the burn-up will be high and the decay cooling time is expected to be short. There is need for liquid–liquid contactors designed for use with these systems and for reflux systems for the ends of cascades when high separations are desired.
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Nuclear Wastes: Technologies for Separations and Transmutation Electrochemical Separations The electrochemical technique known as electrorefining was developed for actinide systems as a simple way to purify plutonium metal alloys proposed as fuel for the Los Alamos Molten Plutonium Reactor Experiment (LAMPRE). The basic separations technique has been adapted to the purification of plutonium from several elements including americium, and to prepare pure uranium and plutonium metals for both the weapons and the breeder-reactor fuels programs (Coops et al., 1983; Christensen and Mullins, 1983). Electrorefining is classified as a pyrochemical process in this instance, because it also uses a molten salt as an ion-transfer medium between an impure metal anode and the pure metal cathode (the collected product). The process somewhat resembles the salt transport process described above, but the driving potential is provided electrically and can be controlled to provide a satisfactory cathode product. In its simplest form, the electrorefining technique is based on the sequential oxidation of the most chemically reactive elements from a molten pool of feed metal (the anode) with transport of the cations through an ionic molten salt to the cathode, where it is reduced to metal and deposited. If the most reactive species present in the anode is an undesirable impurity, the cell potential can be controlled so that ion remains unreduced in the salt phase. In this way, the cell potential can be used to establish a narrow voltage window that permits either a single element to be deposited at the cathode or, in the case of multiple cathodes, different elements to be deposited sequentially and selectively on separate cathodes. When used in the latter mode, the electrorefining process becomes a batch processing technique. Impurity elements with a more negative chemical-free energy than the desired product remain in the transport salt. Elements with more positive free energies remain in the anode pool, either as a solution with the feed or as a sludge when the selected element is depleted from the anode. Argonne National Laboratory Program in Pyroprocessing as Related to Integral Fast Reactor and Light Water Reactor Fuel Recycling Pyroprocessing, as proposed by Argonne National Laboratory (ANL), has the advantages of high density, compact size, and fast kinetics as a consequence of the use of liquid metals, the high concentrations of the elements in molten salts, the temperatures employed, and the generally adequate element-to-element separation factors in the electrochemistry of molten-salt systems. The technologies required are under development and demonstration at present. INTEGRAL FAST REACTOR (IFR) PYROPROCESSING SEPARATIONS The pyrochemical separations required for the integral fast reactor (IFR) pyroprocessing have been demonstrated at the laboratory and bench-engineering levels. These are mainly molten salt electrochemical processes for the metal fuel dissolution and the deposition of the uranium metal onto a solid cathode, with the plutonium and minor TRUs reduced from the molten salt into a liquid cadmium cathode. The technology at ANL appears quite feasible with
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Nuclear Wastes: Technologies for Separations and Transmutation respect to the IFR process. In ANL's program, the spent IFR fuel elements would be maintained in storage for about a year prior to processing to permit short half-like products to decay and eliminate the need for cooling the molten-salt baths during the pyroprocessing steps. Waste gases would be condensed via a cryogenic system involving the argon atmosphere of the working enclosures. They are not considered further in this document; this problem has been addressed by ANL. Since these are mainly elevated temperature processes, the pyrochemical separation processes are dominated by thermodynamic rather than kinetic considerations. This is in contrast to most ambient-temperature electrochemical processes. A process flowsheet is attached as Figure D-1. In the ANL pyroprocess, the minor actinides (americium and curium) follow plutonium. Data provided by ANL demonstrate that, in the process employed, elements of atomic number higher than uranium have similar but different thermodynamic properties. Consequently, large separation factors are observed between uranium and the other TRUs, but the separation factors among the individual TRUs are much smaller. In the ANL pyroprocess, elements are separated in groups according to their inherent thermodynamic properties. The transuranics behave as one group, uranium behaves differently, most of the rare earths behave as another distinct group, and so on. ANL is in the process of defining the various primary and secondary waste streams. The process for IFR pyroprocessing, as it is currently being developed, will result in two new main high-level waste (HLW) forms. The first of these will be a natural or artificial zeolite for the waste salt stream containing the strontium, cesium, and iodide salts and the divalent rare earths that do not reduce readily to the metals and hence are partitioned to the electrorefiner transport salt phase. The second will be a metal matrix of copper, or copper-aluminum, containing the noble metals, some of the rare earths, and perhaps the fuel element undissolved hulls. Both the zeolite and metal matrix waste would constitute new waste forms for the repository. The strontium and cesium fission products, as well as iodide and some of the rare earths, are retained in the KCl/LiCl molten-salt waste stream and are selectively sorbed from the salt by contacting the molten salt with solid zeolites. After hot pressing, a monolithic mineralized material would result (sodalite), which may offer potential as a new waste form. The exact form of the zeolite waste, however, remains to be determined. Preliminary experimental data from work with fine powders of a selected zeolite indicate that the zeolite leach rates can be at least comparable to borosilicate glass. The general concept appears sound with the available data indicating technological feasibility. Regarding separation factors, both chemical models and some laboratory experiments indicate that LLW streams can be partitioned to contain less than 100 nCi/g of TRUs. The uranium that is produced for use as blanket pins has only traces of plutonium remaining in it, and the plutonium/minor actinide fraction appears to contain only small amounts of uranium, so that the driver feed can be formulated as needed. This is not a problem in terms of the fuel for the IFR. Since liquid cadmium is employed in the process, some cadmium would find its way into a melt and zeolite waste streams; ANL is attempting to minimize this. The IFR process depends in part on the thermodynamic driving force generated by the presence of cadmium and the addition of cadmium chloride as an oxidizing agent. Cadmium retorting, recovery, and recycle are essential to the success of the process.
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Nuclear Wastes: Technologies for Separations and Transmutation FIGURE D-1 IFR process flowsheet Source: Argonne National Laboratory - Integral Fast Reactor Program, managed by the University of Chicago for the U.S. Department of Energy under Contract N. W-31-109-Eng-38.
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Nuclear Wastes: Technologies for Separations and Transmutation VOLATILIZATION Volatilization of RuO4 after ozonolysis was discussed earlier. Other volatile species such as the fluorides RuF6 and TcF6 can also be used in separation schemes in which the oxidation states are selectively achieved (Abney, 1991). ELECTRODEPOSITION It may be possible to selectively reduce and deposit technetium in the presence of ruthenium due to the differences in their REDOX potentials. The separations of ruthenium and technetium are probably only relevant in the transmutation scenario. Technetium is reduced to the corrosion-resistant metal by electrodeposition as a possible disposal form. Technetium alloys with the other noble metals should be even more corrosion resistant. MAGNETIC SEPARATION The possibility of magnetically separating (diamagnetic) from reduced forms of ruthenium which are paramagnetic is under consideration. The technique has even been used on solid suspensions in aqueous media (Abney, 1991). Engineering Challenges to Separations HANDLING AND DRYING OF SLUDGES AND SLURRIES The Hanford single-shell tanks typically contain a layer of sludge under a liquid, with a salt cake on top of the liquid. It would be useful to process the sludge, and in any event it would be necessary to move it. Air-lift circulation has been proposed at Hanford for mixing the sludge with wash liquids. If this method presents problems in suspending the sludge, it would be valuable to explore other methods of mixing and resuspension. Another need would be to dewater the sludge. Radioactivity would hamper the use of conventional drying methods. Consequently, novel approaches, such as electroosmotic dewatering, may be applicable. EVAPORATORS Evaporation is already in use at Hanford and can be expected to have a prominent role in processing techniques for nuclear wastes. Fouling of heat transfer surfaces can be expected to be a problem. Because of difficulty of maintenance, it would be important to identify or generate designs for which the fouling problem is avoided or minimized. Vacuum evaporation may be attractive, since lower temperatures can lessen the occurrence of unwanted reactions.
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Nuclear Wastes: Technologies for Separations and Transmutation Various forms of evaporators should be assessed for reliable and maintenance-free operation. Given the nature of wastes, noncondensible gases can be expected to collect in evaporators, and means must be incorporated to purge them so as to avoid substantial lowering of heat-transfer coefficients. EXTRACTORS Centrifugal extractors are effective for reducing equipment volume and residence time. It would be important to locate or identify designs that provide virtually maintenance-free operation for extended periods while avoiding build-up of solids and phases of intermediate density. ENGINEERING OPPORTUNITIES Data from Hanford and ANL show that a potentially useful separation occurs among the three layers in storage tanks. Actinides and some fission products concentrate into the sludge, because the hydroxides are insoluble under alkaline conditions. The sodium salts (nitrate, nitrite, carbonate) compose the salt cake, which probably also incorporates strontium and cesium. Other fission products should partition among the layers. Sludge washing and salt dissolution is proposed as a first step in the processing of the Hanford tanks and should be effective for reduction of sludge volume through removal of nonradioactive, soluble bulk (e.g., sodium salts). Sludge washing, soluble salt purification, and ion exchange could usefully be staged, through implementation of the classical "Shanks-tanks" approach (King, 1980), as used in extraction of instant coffee and other applications. In this scheme, piping is provided such that the position of any one tank within a sequence of tanks is advanced one position at a time, through appropriate opening and closing of valves. One stage of washing, recrystallization, or adsorption could take place for each cycle. This provides the equivalent of countercurrent flow, without actually removing the sludge, ion exchange resin, other solid, or water insoluble liquid from the tank until sludge washing is complete. This approach increases the recovery of salts from the sludge and minimizes the volume of wash water, and the technique might be used for any of the aqueous operations with liquids or solids in the facility. An opportunity should also exist for recrystallizing the salt cake, in order to purify it as much as may be needed to produce LLW. The approach is to redissolve the cake in an aqueous phase and then recrystallize, probably in an evaporative crystallizer. Again, this can be staged, with supernatant liquids transported countercurrent to the salt cake and with recrystallization by heating and cooling at each stage (e.g., by the Shanks system).
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Nuclear Wastes: Technologies for Separations and Transmutation Summary None of the proposed separation methods going beyond the basic ion exchange and PUREX with its auxiliaries in known equipment has been tested sufficiently to be employed immediately. Some are potentially applicable to defense wastes, others to transmutation (partitioning), and some to conventional reactor fuel reprocessing. Of course, some are of possible value in several such areas. A summary of the committee's evaluation of the technologies reviewed is given in Table D-1. The priority ratings reflect a combination of probability of successful development of the technology and the time at which the technology would be useful. For example, TRUEX is rated high based on its likely success and the immediate value of such a technology in processing the defense wastes. By contrast, diphosphine oxides and diamides are given a medium priority, as they would serve the same purpose as TRUEX but have not been as fully evaluated. Talspeak and Tramex are given medium priority ratings as they are not needed at the present time, even though these technologies seem quite promising. Thus, it is important to assess the committee's ratings as reflecting a proposed strategy for developing a program of separations research and development in which the urgency of the need of the technology and its potential for successful development are both reflected.
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Nuclear Wastes: Technologies for Separations and Transmutation TABLE D-1 Summary of Potential Separation Technologies Technology Utilization Development Stage Time of Need Priority Separation of Actinides, Cs, or Sr Stereo-specific Extractants Crown ethers, SREX Defense wastes reprocessing Ready for pilot-plant tests As soon as possible High Pu complexants Defense wastes partitioning Basic lab research Whenever developed High/medium Talspeak Partitioning Advanced lab studies Next decade High/medium Tramex Partitioning Advanced lab studies Next decade High/medium Bidentate Extractants Carbamoyl Phosphine Oxides (TRUEX) Defense wastes partitioning reprocessing Ready for pilot-plant tests As soon as possible High Diamides and diphosphines oxides Defense wastes partitioning reprocessing Advanced lab studies Whenever developed High/medium Molten Salt Reprocessing Ready for pilot-plant tests Next decade High Soft Donor Complexants Defense wastes partitioning Basic lab research Whenever developed High/medium Dicarbollide Complexation Defense wastes partitioning reprocessing Advanced lab studies Whenever developed Medium Super Critical Fluid Chromatography Defense wastes reprocessing Basic lab research Whenever developed Low Organic Resins Defense wastes partitioning reprocessing Advanced lab studies pilot plant As soon as possible High
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Nuclear Wastes: Technologies for Separations and Transmutation Inorganic Exchangers Defense wastes partitioning Basic to advanced lab studies As soon as possible High Adsorption Defense wastes partitioning Basic to advanced lab studies As soon as possible Medium Ultrafiltration Microfiltration Defense wastes Advanced lab studies Whenever developed High Electrolysis Defense wastes Basic lab research Whenever developed High/medium Facilitated Transport Defense wastes Basic lab research Whenever developed Medium Reverse Osmosis Defense wastes Basic lab research Whenever developed Low Dialysis Defense wastes Basic lab research Whenever developed Low Fluorides Reprocessing Advanced lab research As soon as possible High β-Diketones Reprocessing Advanced lab research Next decade Medium Chlorides Reprocessing Advanced lab research Next decade Low Atomic Vapor Partitioning reprocessing Advanced lab research Whenever developed Low Precipitation Defense wastes partitioning Basic lab research As soon as possible Medium Siderophore (microbial) Defense wastes Basic lab research Whenever developed High/medium Jimson Weed Defense wastes Basic lab research Whenever developed Medium Chitin Defense wastes Basic lab research Whenever developed Medium
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Nuclear Wastes: Technologies for Separations and Transmutation Fused Salt Reprocessing Ready for pilot-plant tests Next decade High Magnetic Separation Partitioning Advanced lab research Whenever developed Medium Electromigration, Electrophoresis Partitioning Advanced lab research Whenever developed Low Centrifugation Partitioning reprocessing Basic lab research Next decade Low Separation of Technetium Solvent Extraction, Ion Exchange Defense wastes partitioning Advanced lab research Whenever developed High/medium Ozonolysis Defense wastes partitioning Advanced lab research Whenever developed High/medium Volatilization Defense wastes partitioning Advanced lab research Whenever developed High/medium Electrodeposition Defense wastes partitioning Advanced lab research Whenever developed Medium Magnetic Separation Defense wastes partitioning Advanced lab research Whenever developed Medium/low Engineering Handling and Drying of Sludges and Slurries Defense wastes Pilot-plant tests Whenever developed High Evaporators Defense wastes Pilot-plant tests Whenever developed High Extractors Defense wastes Pilot-plant tests Whenever developed High Engineering Opportunities Defense wastes Pilot-plant tests Whenever developed High
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Nuclear Wastes: Technologies for Separations and Transmutation Yakovlev, G. N., and D. S. Gorbenko-Germanov. 1956. Coprecipitation of americium (v) with double carbonates uranium (vi) or platinum (vi) with potassium. In Proc. 1st U.N. International Conference on Peaceful Uses of Atomic Energy. Vol. 7. Vienna: International Atomic Energy Agency.
Representative terms from entire chapter: