A method of forming a water resistant boundary on a fissile material for use in a water cooled nuclear reactor is described. The method comprises mixing a powdered fissile material selected from the group consisting of UN and U.sub.3Si.sub.2 with an additive selected from oxidation resistant materials having a melting or softening point lower than the sintering temperature of the fissile material, pressing the mixed fissile and additive materials into a pellet, sintering the pellet to a temperature greater than the melting point of the additive. Alternatively, if the melting point of the oxidation resistant particles is greater than the sintering temperature of UN or U.sub.3Si.sub.2, then the oxidation resistant particles can have a particle size distribution less than that of the UN or U.sub.3Si.sub.2

A method of forming a water resistant boundary on a fissile material for use in a water cooled nuclear reactor is described. The method comprises mixing a powdered fissile material selected from the group consisting of UN and U.sub.3Si.sub.2 with an additive selected from oxidation resistant materials having a melting or softening point lower than the sintering temperature of the fissile material, pressing the mixed fissile and additive materials into a pellet, sintering the pellet to a temperature greater than the melting point of the additive. Alternatively, if the melting point of the oxidation resistant particles is greater than the sintering temperature of UN or U.sub.3Si.sub.2, then the oxidation resistant particles can have a particle size distribution less than that of the UN or U.sub.3Si.sub.2.

A method of forming a water resistant boundary on a fissile material for use in a water cooled nuclear reactor is described. The method comprises mixing a powdered fissile material selected from the group consisting of UN and U.sub.3Si.sub.2 with an additive selected from oxidation resistant materials having a melting or softening point lower than the sintering temperature of the fissile material, pressing the mixed fissile and additive materials into a pellet, sintering the pellet to a temperature greater than the melting point of the additive. Alternatively, if the melting point of the oxidation resistant particles is greater than the sintering temperature of UN or U.sub.3Si.sub.2, then the oxidation resistant particles can have a particle size distribution less than that of the UN or U.sub.3Si.sub.2.

A reflector assembly for a molten chloride fast reactor (MCFR) includes a support structure with a substantially cylindrical base plate, a substantially cylindrical top plate, and a plurality of circumferentially spaced ribs extending between the base plate and the top plate. The support structure is configured to encapsulate a reactor core for containing nuclear fuel. The MCFR also includes a plurality of tube members disposed within the support structure and extending axially between the top plate and the bottom plate. The plurality of tube members are configured to hold at least one reflector material to reflect fission born neutrons back to a center of the reactor core.

A reflector assembly for a molten chloride fast reactor (MCFR) includes a support structure with a substantially cylindrical base plate, a substantially cylindrical top plate, and a plurality of circumferentially spaced ribs extending between the base plate and the top plate. The support structure is configured to encapsulate a reactor core for containing nuclear fuel. The MCFR also includes a plurality of tube members disposed within the support structure and extending axially between the top plate and the bottom plate. The plurality of tube members are configured to hold at least one reflector material to reflect fission born neutrons back to a center of the reactor core.

Fission reactor has a cladding encasing a heat generating source including a fissionable nuclear fuel composition. The heat generating source is offset from the surface of the cladding and molten metal is located within the void space formed by the offset. As a liquid, the molten metal will flow and occupy any contiguous network of void space within the fuel cavity and provides thermal transfer contact between the heat generating source and the cladding. The cladding separates the heat generating source and the molten metal from the primary coolant volume.

Fission reactor has a cladding encasing a heat generating source including a fissionable nuclear fuel composition. The heat generating source is offset from the surface of the cladding and molten metal is located within the void space formed by the offset. As a liquid, the molten metal will flow and occupy any contiguous network of void space within the fuel cavity and provides thermal transfer contact between the heat generating source and the cladding. The cladding separates the heat generating source and the molten metal from the primary coolant volume.

There is provided a core of a fast reactor capable of achieving a sodium-cooled metal fuel fast reactor with high adaptability to a molten salt heat storage system, by flattening the output distribution and raising the coolant outlet temperature while suppressing deterioration of the core characteristic. A core of a fast factor is a fuel assembly obtained by densely disposing fuel rods within a wrapper tube, the fuel rod storing, within a cladding tube, hollow fuel in which Pu-enrichment is made to be a predetermined value within a range of 11 to 13 wt %. In the core of a fast factor, a first fuel assembly including a fuel rod with a large hollow diameter of the hollow fuel is loaded on the center side of the core, and a second fuel assembly including a fuel rod with a hollow diameter smaller than the hollow diameter of the hollow fuel of the first fuel assembly is loaded on the circumferential side of the core.

There is provided a core of a fast reactor capable of achieving a sodium-cooled metal fuel fast reactor with high adaptability to a molten salt heat storage system, by flattening the output distribution and raising the coolant outlet temperature while suppressing deterioration of the core characteristic. A core of a fast factor is a fuel assembly obtained by densely disposing fuel rods within a wrapper tube, the fuel rod storing, within a cladding tube, hollow fuel in which Pu-enrichment is made to be a predetermined value within a range of 11 to 13 wt %. In the core of a fast factor, a first fuel assembly including a fuel rod with a large hollow diameter of the hollow fuel is loaded on the center side of the core, and a second fuel assembly including a fuel rod with a hollow diameter smaller than the hollow diameter of the hollow fuel of the first fuel assembly is loaded on the circumferential side of the core.

A nuclear device including a plurality of heat pipes; a first fuel configured to surround respective of the plurality of heat pipes coaxially with respect to a central axis of each of the respective heat pipes, the first fuel containing a fissile material at a first enrichment level; a second fuel configured to directly abut the first fuel on the outside of the first fuel and farther than the first fuel from the respective heat pipes surrounded by the first fuel, the second fuel containing the fissile material at a second enrichment level less than the first enrichment level; and a core including the heat pipes arranged in parallel with each other.