A method for manufacturing a nuclear fuel compact is provided. The method includes forming an additive structure, consolidating a fuel matrix around the additive structure, and thermally processing the fuel matrix to form a fuel compact in which the additive structure is encapsulated therein. The additive structure optionally includes a vertical segment and a plurality of arm segments that extend generally radially from the vertical segment for conducting heat outwardly toward an exterior of the fuel compact. In addition to improving heat transfer, the additive structure may function as burnable absorbers, and may provide fission product trapping.

A device for detecting transport boats includes a contact element for contacting a transport boat, and a connecting element spring mounted in a housing of the device, biased into an initial position and linearly displaceably guided via a guide of the housing. The contact element is connected to the spring mounted connecting element and is displaceable together therewith in such a way that contact of the transport boat with the contact element causes deflection of the connecting element against the bias from the initial position into a detection position. The device further includes a detection device adapted to detect reaching of the detection position by the connecting element.

Various embodiments of the disclosure provide methods using spark plasma sintering (SPS) at moderate temperatures and moderate pressures to fabricate high-density graphite material. The moderate temperatures may be temperatures not exceeding about 1200° C. The moderate pressures may be pressures not exceeding about 300 MPa. The high density exhibited by the resulting, sintered, high-density graphite material may be greater than about 1.75 g/cm.sup.3 (e.g., greater than about 2.0 g/cm.sup.3).

Various embodiments of the disclosure provide methods using spark plasma sintering (SPS) at moderate temperatures and moderate pressures to fabricate high-density graphite material. The moderate temperatures may be temperatures not exceeding about 1200° C. The moderate pressures may be pressures not exceeding about 300 MPa. The high density exhibited by the resulting, sintered, high-density graphite material may be greater than about 1.75 g/cm.sup.3 (e.g., greater than about 2.0 g/cm.sup.3).

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material, the interior surface of which is lined in higher temperature regions with an insulation layer of porous refractory ceramic material. A continuous insulation layer extends the length of the fuel assembly or separate insulation layer sections have a thickness increasing step-wise along the length of the fuel assembly from upper (inlet) section towards bottom (outlet) section. A fuel element positioned inward of the insulation layer and between support meshes has a fuel composition including HALEU and has the form of a plurality of individual elongated fuel bodies or one or more fuel monolith bodies containing coolant flow channels. Fuel assemblies are distributively arranged in a moderator block, with upper end of the outer structural member attached to an inlet for propellant and lower end of the outer structural member operatively interfaced with a nozzle forming a NTP reactor.

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material (e.g., SiC—SiC composite), insulation layer of porous refractory ceramic material (e.g., zirconium carbide with open-cell foam structure or fibrous zirconium carbide), and interior structural member of refractory ceramic-graphite composite material (e.g., zirconium carbide-graphite or niobium carbide-graphite). Spacer structures between various layers provide a defined and controlled spacing relationship. A fuel element bundle positioned between support meshes includes a plurality of distributively arranged fuel elements or a solid, unitary fuel element with coolant channels, each having a fuel composition including high assay, low enriched uranium (HALEU). Fuel assemblies are distributively arranged in a moderator block and the upper end of the outer structural member is attached to a metallic inlet tube for hydrogen propellant and the lower end of the outer structural member is interfaced with a support plate, forming a NTP reactor.

Provided herein is a control method for volume fraction of multistructural isotropic fuel particles in a fully ceramic microencapsulated nuclear fuel including: preparing a mixture of silicon carbide, sintering additives, and organic binders, producing a coating body by coating multistructural isotropic fuel particles by using the prepared mixture, forming the coating body, and performing pressureless sintering on the formed coating body, wherein volume fraction of multistructural isotropic nuclear fuel particles may be controlled by controlling the coating layer thickness on multistructural isotropic nuclear fuel particles, wherein the coating layer was configured with a mixture of silicon carbide, sintering additives, and organic binders. As described above, stability and tolerance against nuclear fuel related accidents may be significantly enhanced, and advantageous effects of enabling a pressureless sintering procedure to be performed while maximizing volume fraction of the multistructural isotropic fuel particles may be expected.

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

Described herein are Uranium silicide materials as advanced nuclear fuel replacements for uranium dioxide fuel in light water reactors (LWRs) that have advantages over currently used uranium dioxide (UO.sub.2) via a substantially higher thermal conductivity and, thus, are capable of operating in a reactor at significantly lower temperatures for the same level of power production, plus the heat capacity of a silicide is lower than that of an oxide so that less heat is stored in the fuel that would need to be removed under accident conditions.