Liquid Fluoride Thorium Reactor - Removal of Fission Products

Removal of Fission Products

Molten salt reactors can benefit greatly from a mechanism to remove the fission products from the fuel salt. Some fission products in the salt absorb neutrons and reduce the production of new fissile fuel. Especially the concentration of some of the rare earth elements needs to be kept low, as they have a large cross section for neutron capture. Some other elements with a small cross section like Cs or Zr can be tolerated in much higher concentrations, so they may accumulate over years of operation.

Removal of fission products is similar to reprocessing of solid fuel elements, without the need to remove and rebuild the fuel cladding. As the fuel of a LFTR is a molten salt mixture, it is attractive to use pyroprocessing, high temperature methods working directly from the hot molten salt. Pyroprocessing does not use radiation sensitive solvents and is not easily disturbed by decay heat. It can be used on the highly radioactive fuel directly from the reactor. Having the chemical separation on site, close to the reactor avoids transport and keeps the total inventory of the fuel cycle low. Ideally everything except new fuel (thorium) and waste (fission products) stays inside the plant.

On site processing is planned to work continuously, cleaning a small fraction of the salt every day and sending it back to the reactor. There is no need to make the fuel salt very clean, the purpose is to keep the concentration of fission products and other impurities (e.g. oxygen) low enough.

The more "noble" metals (Pd, Ru, Ag, Mo, Nb, Sb, Tc... ) do not form fluorides in the normal salt, but form fine metallic particles in the salt. They can plate out at metal surfaces like the heat exchanger or some kind of high surface area filters that are easier to remove. Still there is some uncertainty where these noble elements end up, as the MSRE only provided a relatively short operating experience and independent laboratory experiments are difficult.

Some elements like Xe and Kr come out easily as gas, assisted by a sparge of helium. In addition a part of the "noble" metals are removed together with the gas as a fine mist. Especially the fast removal of Xe-135 is important, as this a very strong neutron poison and makes reactor control more difficult if left in the reactor. Removal of Xe also improves neutron economy. The gas (mainly He, Xe and Kr) is held up for about 2 days until a large fraction of the Xe-135 and other short lived isotopes have decayed. Most of the gas can then be recycled. After an additional hold up of several months, radioactivity is low enough to separate the gas at low temperatures into helium (for reuse), xenon (for sale) and krypton. The krypton needs storage (e.g. in compressed form) for an extended time (several decades) to wait for the decay of Kr-85.

For cleaning the salt mixture several methods of chemical separation were proposed. Compared to classical Purex reprocessing pyroprocessing can be more compact and produce less secondary wastes. The pyroprocesses of the LFTR salt already starts with a suitable liquid form—so it may be cheaper than for solid oxide fuel. However no complete molten salt reprocessing plant was build—just laboratory test of some elements. There is still more research and development needed to improve separation and make reprocessing more economic.

Uranium and some other elements can be removed from the salt by a process called fluorine volatility: A sparge of fluorine removes volatile high-valence fluorides as a gas. This is mainly uranium hexafluoride, containing the uranium-233 fuel, but also neptunium hexafluoride, technetium hexafluoride and selenium hexafluoride, as well as fluorides of various highly radioactive short-lived fission products such as iodine-131, 99molybdenum, and 132tellurium. The volatile fluorides can be further separated by adsorption and distillation. Handling uranium hexafluoride is well established in enrichment. The higher valence fluorides are quite corrosive at high temperatures and require more resistant materials than Hastelloy. One suggestion in the MSBR program at ORNL was using solidified salt as a protective layer. At the MSRE reactor fluorine volatility was used to remove uranium from the fuel salt. Also for use with solid fuel elements fluorine volatility is quite well developed and tested.

Another simple method, tested during the MSRE program, is high temperature vacuum distillation. The lower boiling point fluorides like uranium tetrafluoride and the LiF and BeF carrier salt can be removed by distillation. Under vacuum the temperature can be lower than the ambient pressure boiling point. So a temperature of about 1000 °C is sufficient to recover most of the FLiBe carrier salt. However, while possible in principle, separation of thorium fluoride from the even higher boiling point lanthanide fluorides would require very high temperatures and new materials. The chemical separation for the 2-fluid designs, using uranium as a fissile fuel can work with these two relatively simple processes: Uranium from the blanket salt can be removed by fluorine volatility, and transferred to the core salt. To remove the fissile products from the core salt, first the uranium is removed via fluorine volatility. Then the carrier salt can be recovered by high temperature distillation. The fluorides with a high boiling point, including the lanthanides stay behind as waste.

The early Oak Ridge's chemistry designs were not concerned with proliferation and aimed for fast breeding. They planned to separate and store protactinium-233, so it could decay to uranium-233 without being destroyed by neutron capture in the reactor. With a half-life of 27 days, 2 months of storage would assure that 75% of the 233Pa decays to 233U fuel. The protactinium removal step is not required per se for a LFTR. Alternate solutions are operating at a lower power density and thus a larger fissile inventory (for 1 or 1.5 fluid) or a larger blanket (for 2 fluid). Also a harder neutron spectrum helps to achieve acceptable breeding without protactinium isolation.

If Pa separation is specified, this must be done quite often (for example, every 10 days) to be effective. For a 1 GW, 1-fluid plant this means about 10% of the fuel or about 15 t of fuel salt need to go through reprocessing every day. This is only feasible if the costs are much lower than current costs for reprocessing solid fuel.

Newer designs usually avoid the Pa removal and send less salt to reprocessing, which reduces the required size and costs for the chemical separation. It also avoids proliferation concerns due to high purity U-233 that might be available from the decay of the chemical separated Pa.

Separation is more difficult if the fission products are mixed with thorium, because thorium, plutonium and the lanthanides (rare earth elements) are chemically similar. One process suggested for both separation of protactinium and the removal of the lanthanides is the contact with molten bismuth. In a redox-reaction some metals can be transferred to the bismuth melt in exchange for lithium added to the bismuth melt. At low lithium concentrations U, Pu and Pa move to the bismuth melt. At more reducing conditions (more lithium in the bismuth melt) the lanthanides and thorium transfer to the bismuth melt too. The fission products are then removed from the bismuth alloy in a separate step, e.g. by contact to a LiCl melt. However this method is far less developed. A similar method may also possible be with other liquid metals like aluminum.