The present disclosure relates to nuclear chemical, particularly radiochemical, technologies at different stages of the nuclear fuel cycle, such as the production of purified nuclear materials (uranium, zirconium) or the reprocessing of spent nuclear fuel from nuclear power stations, in which extraction processes and operations for purifying nuclear materials are used. An example method, which includes the partial decomposition of nitric acid during continuous evaporation while a solution containing a reducing agent is fed into the bottom part of an evaporator having a circulating bottoms solution, consists in that the process is carried out such that the solution is kept in the bottom part of the evaporator for more than 2 hours under the addition of an aqueous solution of formaldehyde and formic acid (hereinafter “the mixture”) or a solution of formic acid after the process has been started using the mixture.

The present disclosure relates to nuclear chemical, particularly radiochemical, technologies at different stages of the nuclear fuel cycle, such as the production of purified nuclear materials (uranium, zirconium) or the reprocessing of spent nuclear fuel from nuclear power stations, in which extraction processes and operations for purifying nuclear materials are used. An example method, which includes the partial decomposition of nitric acid during continuous evaporation while a solution containing a reducing agent is fed into the bottom part of an evaporator having a circulating bottoms solution, consists in that the process is carried out such that the solution is kept in the bottom part of the evaporator for more than 2 hours under the addition of an aqueous solution of formaldehyde and formic acid (hereinafter “the mixture”) or a solution of formic acid after the process has been started using the mixture.

The invention relates to technology for processing liquid radioactive waste containing, inter alia, tritium isotopes, which are formed in various nuclear industry plants, and also during decommissioning of such plants. The technical result of the claimed invention consists in simplifying the technological procedure for processing liquid radioactive waste containing, inter alia, tritium isotopes by excluding complicated and lengthy operations associated with testing a concrete mixture produced from deactivated liquid radioactive waste, and also in increasing the ecological safety by reducing the size of areas for storage of the waste produced during the processing of the liquid radioactive waste. The claimed technical result is achieved in that a method for processing liquid radioactive waste containing, inter alia, tritium isotopes involves removing radioactive substances from the liquid radioactive waste so as to produce a low-level waste solution, and introducing a binder into the low-level waste solution produced in order to prepare a concrete mixture which complies with structural, radioecological, and sanitary and hygiene requirements, wherein components that have a negative effect on the technical characteristics of the concrete mixture being produced are removed from the low-level waste solution before the binder is added.

The invention relates to technology for processing liquid radioactive waste containing, inter alia, tritium isotopes, which are formed in various nuclear industry plants, and also during decommissioning of such plants. The technical result of the claimed invention consists in simplifying the technological procedure for processing liquid radioactive waste containing, inter alia, tritium isotopes by excluding complicated and lengthy operations associated with testing a concrete mixture produced from deactivated liquid radioactive waste, and also in increasing the ecological safety by reducing the size of areas for storage of the waste produced during the processing of the liquid radioactive waste. The claimed technical result is achieved in that a method for processing liquid radioactive waste containing, inter alia, tritium isotopes involves removing radioactive substances from the liquid radioactive waste so as to produce a low-level waste solution, and introducing a binder into the low-level waste solution produced in order to prepare a concrete mixture which complies with structural, radioecological, and sanitary and hygiene requirements, wherein components that have a negative effect on the technical characteristics of the concrete mixture being produced are removed from the low-level waste solution before the binder is added.

This invention has an object to provide a method for producing a beryllium solution, the method being novel and having high energy efficiency. The method (M10) for producing a beryllium solution includes a main heating step (S13) of dielectrically heating an acidic solution containing a starting material so as to generate a beryllium solution, the starting material being beryllium or a substance containing beryllium.

The invention concerns an apparatus and a method for treating radioactive material (36), in particular for cleaning radioactive contaminated water. The apparatus comprises a process chamber (10) with a combustion zone (12) for generating an oxygen rich gas (34) and an oxidation zone (14), which is arranged to receive the oxygen rich gas (34) from the combustion zone (12). The process chamber (10) further comprises a feed opening (16) for feeding the radioactive material (36) into the oxidation zone (14) and the process chamber (10) is configured to use the oxygen rich gas (34) for oxidizing the radioactive material (36) to obtain oxidized material (38). The apparatus further comprises a separation device (50) operationally connected to an outlet of the process chamber (10) and configured to at least partly separate the oxidized material (38) into a gaseous fluid (56) and a non.sup.− gaseous residue (58). This way a greatly reduced volume of the radioactive material (36) is achieved, enabling safe and efficient handling and/or compact and space-saving disposal of the radioactive material (36).

An object to be decontaminated contaminated with radioactive material is introduced into an acid eluting solvent to dissolve the radioactive material. The radioactive material dissolved is concentrated and separated from the eluting solvent in the present method. The object to be decontaminated comprises contaminated soil and contaminated liquid. One or both of the contaminated soil and the contaminated liquid are collected and introduced into the eluting solvent. The radioactive materials and the object to be decontaminated dissolved in the eluting solvent are separated into solid and liquid. The decontaminated soil separated from the eluting solvent is collected. The eluting solvent used for separating the radioactive material and in which radioactive material is dissolved is concentrated.

An object to be decontaminated contaminated with radioactive material is introduced into an acid eluting solvent to dissolve the radioactive material. The radioactive material dissolved is concentrated and separated from the eluting solvent in the present method. The object to be decontaminated comprises contaminated soil and contaminated liquid. One or both of the contaminated soil and the contaminated liquid are collected and introduced into the eluting solvent. The radioactive materials and the object to be decontaminated dissolved in the eluting solvent are separated into solid and liquid. The decontaminated soil separated from the eluting solvent is collected. The eluting solvent used for separating the radioactive material and in which radioactive material is dissolved is concentrated.

A neutron absorber synthesis system that can synthesize boron carbide that is a raw material for a neutron absorber, by recycling boron (B-10) of a mass number 10 that can absorb boron, particularly neutrons existing in boric acid waste fluid, is provided. The neutron absorber synthesis system includes: a pre-processing unit to which a radioactive waste including boron is supplied from the outside and inflows to the inside and a compound is produced by removing moisture of the radioactive waste by heat treatment by a first heat source; and a boron carbide synthesizing unit to which the compound produced from the radioactive waste is inflowed inside and a boron carbide is synthesized from a raw material containing the compound and carbon by heat treatment by a second heat source.

A transport or storage container holding radioactive waste and a body of water is dried by the steps of first draining or pumping out the body of water and thereby leaving residual water in the container. Then at least one solid drying agent is introduced into an interior the container for removing from the interior of the container for removing the physically or chemically bonded residual water. The solid drying agent is an alkaline earth salt, particularly an alkaline earth oxide.