A system and methods for the extraction, solidification and degasification of carbon-14 isotopes (.sup.14C) and other radioactive isotopes from spent ion exchange resins, the system comprising: a washing vessel for washing the spent resins; a reaction vessel for treating the washed spent resins with alkaline chemicals; and a decantation vessel for receiving the reacted solution and the suspended precipitate, thereby allowing for the removal of a supernatant from the remaining solution to form a solid form of carbonate and/or bicarbonate ions product, or salts thereof, comprising the radioactive isotope. A method for the extraction of a radioactive isotope from spent resins of a nuclear plant using alkaline chemicals is also provided. A method for the solidification and degassing of a radioactive isotope from spent resins of a nuclear plant is further provided.

Provided is a fission product processing method for selectively transmuting only a long-lived radionuclide from fission products. The method for processing radioactive waste includes the step of extracting, from the radioactive waste, the isotopes without isotope separation, the isotope elements including radionuclides of fission products and having a common atomic number, and the step of irradiating the isotopes with high-energy particles generated by an accelerator to produce nuclear transmutation of a long-lived radionuclide of the radionuclides into a short-lived radionuclide with a short half-life or a stable nuclide re-utilizable as a resource.

An electrolytic enrichment method for heavy water includes enriching heavy water by electrolysis using an alkaline water electrolysis cell including an anode chamber that holds an anode, a cathode chamber that holds a cathode, and a diaphragm. In the method, an electrolyte prepared by adding high-concentration alkaline water to raw material water containing heavy water is circularly supplied to the anode chamber and the cathode chamber from a circulation tank; an anode-side gas-liquid separator and an anode-side water-seal device are connected to the anode chamber, and a cathode-side gas-liquid separator and a cathode-side water-seal device are connected to the cathode chamber; and electrolysis is continued while the alkali concentration in the electrolyte supplied to both electrolysis chambers is maintained at a constant concentration by circularly supplying, to the circulation tank, the electrolyte from which the gas generated from the anode-side gas-liquid separator and the cathode-side gas-liquid separator is separated.

Systems and methods for producing acid deficient uranyl nitrate from a dilute uranyl nitrate solution are disclosed. In one form, the present disclosure provides a system comprising a feed evaporation system and a diffusion dialysis system. The feed evaporation system is configured to receive a feed stream and to boil off water, under vacuum, from the feed stream to produce a concentrated uranyl nitrate solution and a distilled water product. The diffusion dialysis system is configured to counter flow the concentrated uranyl nitrate solution and the distilled water product across a plurality of membrane vessels to promote nitrate migration from the concentrated uranyl nitrate solution to the distilled water, and to produce a dialysate stream and a recycle acid stream. The feed stream may include a product of a solvent extraction process used to recycle spent nuclear fuel and/or a recovery stream from other fuel fabrication activities.

The present invention relates to a method for treating tritium water-containing raw water, the method including supplying a part of raw water containing tritium water and alkali water to a circulation tank, mixing the raw water with alkali water in the circulation tank to obtain an electrolyte adjusted so as to have a desired alkali concentration, and continuously electrolyzing the electrolyte while circulating the electrolyte, thereby subjecting the raw water stored in the storage tank to alkali water electrolysis and thus gasifying the raw water. According to the invention, by gasifying tritium water-containing raw water by alkali water electrolysis, the tritium concentration in a tritium-containing hydrogen gas is diluted to 1/1,244 and the tritium water-containing raw water can be reduced in volume.

Submersible media filters and submersible columns for use in removing radioactive isotopes and other contaminants from a fluid stream, such as a fluid stream from the primary coolant loop of a nuclear reactor system or a fluid stream from a spent-fuel pool. Generally, these submersible media filters and submersible columns are adapted to be submersed in the fluid stream, and additionally the filters are adapted to be vitrified after use, resulting in a stabilized, non-leaching final waste product with a substantially reduced volume compared to the original filter. In several embodiments, the submersible media filters and submersible columns include isotope-specific media (ISM).

The present invention relates to a method for treating tritium water-containing raw water, the method including supplying a part of raw water containing tritium water and alkali water to a circulation tank, mixing the raw water with alkali water in the circulation tank to obtain an electrolyte adjusted so as to have a desired alkali concentration, and continuously electrolyzing the electrolyte while circulating the electrolyte, thereby subjecting the raw water stored in the storage tank to alkali water electrolysis and thus gasifying the raw water.

According to the invention, by gasifying tritium water-containing raw water by alkali water electrolysis, the tritium concentration in a tritium-containing hydrogen gas is diluted to 1/1,244 and the tritium water-containing raw water can be reduced in volume.

Nuclear reactors produce large amounts of spent fuel during operation. In addition to recyclable materials such as uranium and plutonium, spent fuel also contains significant amounts of fission products and highly radioactive transuranic (TRU) elements. Homogenization of nuclides in traditional post-processing technology hinders efficient fuel recycling. In order to improve the efficiency of spent fuel recycling, and to reduce the inevitable highly radioactive waste produced in existing spent fuel recycling processes, a novel method for pre-separating nuclides in spent fuel is disclosed. Specifically, we have developed a physical method, wherein an artificially created large temperature gradient drives the migration of fission gas bubbles in spent metallic nuclear reactor fuel. The fission gas bubbles preferentially carry fission products and transuranic elements to achieve effective pre-separation of these elements from spent fuel, lowering cost and improving efficiency of spent fuel recycling.

A single integrated system for recycling used nuclear fuel (UNF) emerging from a reactor has a decladding vessel separating fuel pellets from nuclear fuel rods via oxidation to produce U.sub.3O.sub.8. A fluorination vessel is coupled to the decladding vessel to remove hexafluorides from the U.sub.3O.sub.8 produced by the decladding vessel. An electrowinning vessel is coupled to the fluorination vessel removing plutonium and actinides via electrowinning.

Methods for isolating Pb and/or Pb isotopes from various sources are provided. Compositions comprising Pb and/or Pb isotopes free of certain amounts of various contaminants are also provided.