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Molten Salt Chemistry Plays a Prominant Role in Accelerator-Driven Transmutation Systems

The nuclear system being studied by the Accelerator-Driven Transmutation Technology (ADTT) project is an accelerator-driven, subcritical nuclear reactor with molten salt used as the working fluid. Such a system could be used to destroy weapon plutonium and plutonium contained in Defense Program (DP) wastes and spent nuclear fuel. Briefly, a proton beam generated by an accelerator is directed at a target material of high atomic number (e.g., Pb, Bi) resulting in the production of neutrons by spallation. These neutrons are used to cause the plutonium to fission and to enhance the destruction of fission products. The plutonium, present as plutonium trifluoride dissolved in 2 7LiF - BeF2 fuel salt at 600 °C-750 °C, is circulated through the core of the system, heat exchangers, and fission-product-removal modules. The system shares some common characteristics of the reactor designs developed and studied at Oak Ridge National Laboratory (ORNL) during the 1960s and 1970s. The fuel-preparation process and the removal of fission products are essential elements in the operation of these systems. The following paragraphs describe an overview of the chemistry associated with the conceptual fuel-preparation process and the conceptual separation process for the transtion metal fission products, lanthanide fission products, and gaseous fission products.

The essential step in fuel preparation consists of the hydrofluorination of plutonium in the presence of excess hydrogen to produce plutonium trifluoride dissolved in the fuel salt. Excess hydrogen is used to prevent the formation of volatile plutonium hexafluoride. Conversion of weapon plutonium to ADTT fuel requires no preprocessing. However, converting the plutonium contained in DP wastes and spent nuclear fuel to ADTT fuel requires additional processing, as described below.

Treatment of DP wastes consists of the addition of the waste form (e.g., contamin- ated graphite or ceramic) to a secondary 7LiF-NaF-KF molten salt, where the plutonium is hydrofluorinated in the presence of excess hydrogen. The waste matrix (e.g., graphite or ceramic) is removed from the secondary salt by physical means, such as filtration. The secondary salt is used in the hydrofluorination step instead of the fuel salt so that when the waste matrix is removed from the salt, it is not contaminated with BeF2 and may be discarded as nonhazardous waste. Finally, the secondary salt is blended with the fuel salt before it is transferred to the nuclear system.

Figure 4: Accelerator-driven transmutation of waste.

Spent nuclear fuel consists of plutonium dioxide, uranium dioxide, fission products, and metal cladding. The spent fuel is decladded (although decladding is not essential), and the oxide fuel matrix is crushed, added to the fuel salt, and hydrofluorinated to produce soluble fluorides. Excess hydrogen prevents the formation of volatile fluoride species (e.g., MoF6, UF6, PuF6). Gaseous fission products (e.g., Xe, Kr) released during the process must be adsorbed on activated charcoal or molecular sieves. Next, an electrowinning process is used to remove the transition metal fission products and uranium from the fuel salt. The plutonium remains in the fuel salt because its reduction potential falls below that for beryllium, a major constituent of the fuel salt. The electrowinning process is described in more detail later in the article. Finally, the fluoride potential of the fuel salt is adjusted before it is transferred to the nuclear system.

After the fuel salt has been introduced into the nuclear system and as the fission process occurs in the system, the fission product concentration increases in the fuel salt. A class of transition metal fission products that poses several problems is the so-called "noble metal" fission products: Mo, Nb, Ru, Rh, Ag, Cd, etc. These could potentially become plated onto surfaces in the heat exchanger and elsewhere in the system. They also affect the neutron economy of the system. The on-line removal of the noble metals from the molten salt is crucial to the operation of the system. Electrowinning is envisioned for removing the noble metal fission products.

Electrowinning methods have been used extensively to produce pure metals from oxide or halide feed material dissolved in molten salt (e.g., Al, Mg production). In this application of the method, one is interested in purifying the molten salt and not in producing a pure metal. The electrowinning cell consists of a consumable Be anode and a nickel cathode onto which the metals are deposited. The reaction is spontaneous because of the difference in free energy between metal fluorides. Therefore, in principle, the cell could be operated passively (i.e., no external voltage source required). However, the efficiency of the cell would be enhanced by applying an externally generated voltage. Plutonium, higher actinides, lanthanides, Sr, and Cs remain in the fuel salt. The cell could be located ahead of the heat exchanger system to prevent the deposition of fission products on the heat exchanger tubes. Such deposition poses no safety threat to the system in terms of material interactions, but it would degrade the efficiency of the heat transfer system. Maintenance of the cell would consist of the periodic, remote replacement of the anode and cathode. The metals collected on the cathode could be oxidized, blended with silica, and vitrified.

Another class of fission products that must be removed from the fuel salt is the lanthanides. These must be removed to maintain the neutron economy in the system. Also, they must be removed because of the similarity of their solubility and physical properties with trivalent plutonium. The ADTT system has two options for the remov-al of the lanthanide,s and each is being evaluated on the basis of waste minimization and its resistance to plutonium proliferation. The options are described below.

One option for removing the lanthanides is liquid-metal extraction coupled to a liquid-metal centrifuge system. Scientists at ORNL studied the extraction of lanthanides and plutonium from fluoride molten salt using liquid bismuth that contained a small concen-tration of lithium. In this application, the liquid-metal extraction step is used as a feed for the centrifuge system. The lithium concentration is chosen so that both the lanthanides and plutonium are extracted into the liquid bismuth. The bismuth is used to transport the species to the centrifuge system, where they are separated from each other. Plutonium contained in one of the bismuth streams that comes from the centrifuge system could be back-extracted into the fuel salt and fed back into the nuclear system. Lanthanides contained in the other bismuth stream could be removed from that stream by decreasing the temperature of the bismuth, collecting the solids that are formed, blending the solids with silica, and vitrifying the waste material. The bismuth from each stream could be recycled. The centrifuges could be operated at modest speeds and moderate temperatures while still providing adequate separation.

Alternatively, the lanthanides and plutonium could be separated by a multiple-stage, liquid-metal extraction process. In this process, the lithium concentration in the bismuth is selected so that the plutonium is preferentially extracted from the fuel salt in the first extraction stages. A different lithium concentration is selected for use in subsequent extraction stages so that the lanthanides are removed from the fuel salt. The two bismuth streams produced by this option could be processed in the same way as the two streams that come from the centrifuge system.

Finally, xenon and krypton isotopes produced during the fission process also affect the neutron economy of the system and must be removed from the fuel salt. A helium sparge gas could be used to remove the xenon and krypton from the fuel salt and to transport the gases to a fractional distillation system, where they would be separated from the helium and from each other. A similar system was proposed and designed by ORNL for use with the molten salt breeder reactor.

Several of the main chemical processes associated with the fluid-fueled system have been described in this article. Other significant issues are being studied, but these will be left for future discussion. As an ending note, scientists at ORNL have described the molten salt nuclear system as "a chemist's reactor," and it truly is.

Participants in this phase of the ADTT project include N. Li, F. Venneri, and M. Williamson.

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