A structure—for use in simulating radioactive contamination environments—comprises fragments encapsulated within a substrate material. The fragments comprise radioactive isotopes with moderate half-lives. To form such structures, the fragments are encapsulated within the at least one substrate material. In a method of simulating a radioactive contamination environment, multiple removable structures, such as the aforementioned structures, are selectively placed in a facility, and may be subsequently removed, stored, and reused.

A structure—for use in simulating radioactive contamination environments—comprises fragments encapsulated within a substrate material. The fragments comprise radioactive isotopes with moderate half-lives. To form such structures, the fragments are encapsulated within the at least one substrate material. In a method of simulating a radioactive contamination environment, multiple removable structures, such as the aforementioned structures, are selectively placed in a facility, and may be subsequently removed, stored, and reused.

The present invention relates to a method for at least partially digesting a uranium (U)-based target material which comprises at least one uranium-metal (U-Me) alloy containing Mn, Fe, Co or Ni and comprising at least a U6Me phase. By means of an accelerant in a basic solution, the uranium in U-Me alloy oxidizes to U(VI). The accelerant comprises in particular KMnO4 whilst the U-Me alloy comprises a U—Mn alloy. The alloy preferably comprises two phases of an eutectic system, in particular the U6Mn/UMn2 system. The use of the accelerant enables an enhanced digestion of the U-Me alloy.

The present invention relates to a method for at least partially digesting a uranium (U)-based target material which comprises at least one uranium-metal (U-Me) alloy containing Mn, Fe, Co or Ni and comprising at least a U6Me phase. By means of an accelerant in a basic solution, the uranium in U-Me alloy oxidizes to U(VI). The accelerant comprises in particular KMnO4 whilst the U-Me alloy comprises a U—Mn alloy. The alloy preferably comprises two phases of an eutectic system, in particular the U6Mn/UMn2 system. The use of the accelerant enables an enhanced digestion of the U-Me alloy.

Methods and systems for stabilizing spent fuel assemblies from sodium-cooled nuclear reactors using Zamak are described herein. It has been determined that there is a synergism between Zamak and sodium that allows Zamak to form thermally-conductive interface with the sodium-wetted surfaces of the fuel assemblies. In the method, one or more spent fuel assemblies are removed from the sodium coolant pool and placed in a protective sheath. The remaining volume of the sheath is then filled with liquid Zamak. To a certain extent Zamak will dissolve and alloy with sodium remaining on the fuel assemblies. Excess sodium that remains undissolved is displaced from the sheath by the Zamak fill. The Zamak is then cooled until solid and the sheath sealed. The resulting Zamak-stabilized spent fuel assembly is calculated to have sufficient internal thermal conductivity to allow it to be stored and transported without the need for liquid cooling.

Methods and systems for stabilizing spent fuel assemblies from sodium-cooled nuclear reactors using Zamak are described herein. It has been determined that there is a synergism between Zamak and sodium that allows Zamak to form thermally-conductive interface with the sodium-wetted surfaces of the fuel assemblies. In the method, one or more spent fuel assemblies are removed from the sodium coolant pool and placed in a protective sheath. The remaining volume of the sheath is then filled with liquid Zamak. To a certain extent Zamak will dissolve and alloy with sodium remaining on the fuel assemblies. Excess sodium that remains undissolved is displaced from the sheath by the Zamak fill. The Zamak is then cooled until solid and the sheath sealed. The resulting Zamak-stabilized spent fuel assembly is calculated to have sufficient internal thermal conductivity to allow it to be stored and transported without the need for liquid cooling.

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.

A method for reducing radiologically-contaminated waste is provided. The method comprises treating radiologically-contaminated surfaces and subsurfaces. The method comprises consolidating soil waste. The method comprises employing real-time scanning technology to classify waste based at least in part on a threshold of radiological contamination. The waste is sorted based on the classification. The waste is disposed of via at least one of different disposal routes, based at least in part on the classification.

Uranium-chelating polypeptides comprising at least one helix-loop-helix calcium-binding (EF-hand) motif which comprises a deletion of at least two amino acid in the 12-amino-acid calcium-binding loop sequence, and their use for uranium biodetection and biodecontamination.

A purification method that uses ion-selective ceramics to electrochemically filter waste products from a molten salt. The electrochemical method uses ion-conducting ceramics that are selective for the molten salt cations desired in the final purified melt, and selective against any contaminant ions. The method can be integrated into a slightly modified version of the electrochemical framework currently used in pyroprocessing of nuclear wastes.