Abstract: The invention relates to sealing a fuel rod composite cladding tube composed of silicon carbide regardless of the fuel rod cladding design architecture (e.g., monolithic, duplex with monolithic SiC on the inside and a composite made with SiC fibers and SiC matrix on the outside) preferably with sealed SiC end plug caps, additionally sealed with an interior braze and exterior SiC final coating, thus providing a double sealed end plug barrier effective at retaining gas tightness and providing mechanical strength for the sealed end joint while providing high chemical resistance.

Abstract: An improved nuclear fuel that has enhanced oxidation resistance and a process for making it are disclosed. The fuel comprises a composite of U235 enriched U3Si2 particles and an amount less than 30% by weight of UO2 particles positioned along the surface of the U3Si2 particles. The composite may be compressed into a pellet form. The process comprises forming a layer of UO2 on the surface of U3Si2 particles, either by exposing U3Si2 particles to an atmosphere of up to 15% oxygen by volume dispersed in an inert gas for a period of time and at a temperature sufficient to form UO2 at the U3Si2 particle surface, or by mixing U3Si2 particles with an amount up to 30% by weight of UO2 particles.

Abstract: The invention relates to compositions and methods for coating a zirconium alloy (e.g., ceramic-containing) cladding tube for use with fuel rods in a nuclear water reactor. The coating includes an intermediate oxidation resistant layer and a SiC containing layer at least partially deposited on the intermediate oxidation resistant layer. The SiC containing layer can include a plurality of fibers. The SiC containing layer may also be on the outer surface of end plugs. The invention provides improved capability for the zirconium alloy cladding to withstand normal and accident conditions in the nuclear water reactor.

Abstract: A method is provided for coating the substrate of a component, such as a zirconium alloy cladding tube, for use in a water cooled nuclear reactor under normal operating conditions and under high temperature oxidation conditions. The method includes heating a pressurized carrier gas to a temperature between 200° C. and 1200° C., adding chromium or chromium-based alloy particles having an average diameter of 20 microns or less to the heated carrier gas, and spraying the carrier gas and particles onto the substrate at a velocity, preferably from 800 to 4000 ft./sec. (about 243.84 to 1219.20 meters/sec.), to form a chromium and/or chromium-based alloy coating on the substrate to a desired thickness.

Abstract: The invention relates to compositions and methods for coating a zirconium alloy, e.g., ceramic-containing, cladding tube for use with fuel rods in a nuclear water reactor. The coating includes an intermediate oxidation resistant layer and a SiC containing layer at least partially deposited on the intermediate oxidation resistant layer. The SiC containing layer can include a plurality of fibers. The invention provides improved capability for the zirconium alloy cladding to withstand normal and accident conditions to which it is exposed in the nuclear water reactor.

Abstract: A zirconium alloy cladding tube for use in a water cooled nuclear reactor under normal operating conditions and under high temperature oxidation conditions is described. The cladding tube has a coating uniformly deposited thereon. The coating, which may be up to 300 microns thick, is selected from the group consisting of chromium, a chromium-based alloy, and combinations thereof.

Abstract: The invention relates generally to uranium fuel in a nuclear reactor and, more particularly, the inclusion of a fuel additive component to the bulk fuel material. The fuel additive component is selected and provided in an amount such that it is effective to improve one or more properties of the bulk fuel material. The fuel additive component has a grain size that is less than the grain size of the bulk fuel material. The granular fuel additive component coats or covers the granular bulk fuel material.

Abstract: A method for making an improved nuclear fuel cladding tube includes reinforcing a Zr alloy tube by first winding or braiding ceramic yarn directly around the tube to form a ceramic covering, then physically bonding the ceramic covering to the tube by applying a first coating selected from the group consisting of Nb, Nb alloy, Nb oxide, Cr, Cr oxide, Cr alloy, or combinations thereof, by one of a thermal deposition process or a physical deposition process to provide structural support member for the Zr tube, and optionally applying a second coating and optionally applying a third coating by one of a thermal deposition process or a physical deposition process. If the tube softens at 800° C.-1000° C., the structural support tube will reinforce the Zr alloy tube against ballooning and bursting, thereby preventing the release of fission products to the reactor coolant.

Abstract: An improved, accident tolerant fuel for use in light water and lead fast reactors is described. The fuel includes a ceramic cladding, such as a multi-layered silicon carbide cladding, and fuel pellets formed from U15N and from 100 to 10000 ppm of a boron-containing integral fuel burnable absorber, such as UB2 or ZrB2.

Abstract: A molten salt reactor includes a containment vessel, a reactor core, a neutron reflector spaced from the containment vessel, and liquid fuel enclosed within the core. The liquid fuel is comprised of a nuclear fission material dissolved in a molten salt. A heat exchanger is positioned external to the containment vessel. A plurality of heat transfer pipes are provided for transferring heat from the core to the heat exchanger. Each pipe has a first and a second end. The first end of each pipe is positioned within the reactor core for absorbing heat from the fuel. The heat exchanger receives the second end of each heat transfer pipe. At least two or more reactor shut down systems are provided. At least one shut down system may be a passive system and at least one or both shut down systems may be an active or a manually operated system.

Abstract: The method described herein may be characterized as reacting uranium dioxide with carbon to produce uranium carbide, and, reacting the uranium carbide with a silane, a silicon halide, a siloxane, or combinations thereof, and excess hydrogen to produce uranium silicide.

Abstract: A method for making a fuel rod cladding tube and a cladding tube are described. The method includes wrapping ceramic fibers, for example, SiC fibers in a SiC matrix, around a tube formed from a metal alloy, such as a zirconium alloy. The interstices of the SiC wrappings on the tube are at least partially filled with SiC nano-sized particles. The surface of the filled tube is exposed by atomic layer deposition, at temperatures ranging from 25° C. to 600° C., to at least one cycle of alternating, non-overlapping pulses of gaseous precursors containing carbon and silicon to form a SiC monolayer. The step of filling the interstices of the SiC wrappings on the tube with SiC nano-sized particles fills large voids in the SiC wrapping. The step of exposing the surface of the particle filled SiC windings to at least one cycle of gaseous pulses fills small voids in the SiC wrapping.

Abstract: 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 U3Si2 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 U3Si2, then the oxidation resistant particles can have a particle size distribution less than that of the UN or U3Si2.

Abstract: A method of making a nuclear fuel pellet for a nuclear power reactor. The method includes: providing a nuclear fuel material in powder form, the nuclear material is based on UO2; providing an additive; forming a green pellet, wherein said additive is added either to said nuclear fuel material or to the green pellet; and sintering the green pellet, wherein said additive causes larger grains in the nuclear fuel pellet, and wherein said additive is made of or includes a substance which causes the larger grains and which substantially leaves at least an outer portion of the pellet before and/or during the sintering step, wherein said substance is made of, or comprises, B and/or Cr.

Abstract: A zirconium alloy nuclear reactor cylindrical cladding has an inner Zr substrate surface, an outer volume of protective material, and an integrated middle volume of zirconium oxide, zirconium and protective material, where the protective material is applied by impaction at a velocity greater than 340 meters/second to provide the integrated middle volume resulting in structural integrity for the cladding.

Abstract: A zirconium alloy nuclear reactor cylindrical cladding has an inner Zr substrate surface (10), an outer volume of protective material (22), and an integrated middle volume (20) of zirconium oxide, zirconium and protective material, where the protective material is applied by impaction at a velocity greater than 340 meters/second to provide the integrated middle volume (20) resulting in structural integrity for the cladding.

Abstract: An improved, accident tolerant fuel for use in light water and heavy water reactors is described. The fuel includes a zirconium alloy cladding having a chromium or chromium alloy coating and an optional interlayer of molybdenum, tantalum, tungsten, and niobium between the zirconium alloy cladding and the coating, and fuel pellets formed from U3Si2 or UN and from 100 to 10000 ppm of a boron-containing integral fuel burnable absorber, such as UB2 or ZrB2, either intermixed within the fuel pellet or coated over the surface of the fuel pellet.

Abstract: The method described herein uses a ceramic precursors and controlled temperature rises for forming a stiffened ceramic composite fiber matrix to form a ceramic composite fuel cladding tube of the desired geometry and for removing a mandrel about which the composite fiber matrix was formed. The method described herein allows the manufacture of elongated ceramic composite claddings where the mandrel used to define the geometry of the cladding is easily removed without damaging the ceramic composite cladding. The method includes covering ceramic fibers with a mixture comprising at least one precursor of a ceramic material, wrapping the precursor covered fibers around a mandrel, heating the precursor covered fibers to the decomposition temperature of the precursor to convert the precursor to the ceramic material, and heating the mandrel to at least the melting point thereof to remove the mandrel.

Abstract: An improved, accident tolerant fuel for use in light water and lead fast reactors is described. The fuel includes a ceramic cladding, such as a multi-layered silicon carbide cladding, and fuel pellets formed from U15N and from 100 to 10000 ppm of a boron-containing integral fuel burnable absorber, such as UB2 or ZrB2.

Abstract: An improved nuclear fuel that has enhanced oxidation resistance and a process for making it are disclosed. The fuel comprises a composite of U235 enriched U3Si2 particles and an amount less than 30% by weight of UO2 particles positioned along the surface of the U3Si2 particles. The composite may be compressed into a pellet form. The process comprises forming a layer of UO2 on the surface of U3Si2 particles, either by exposing U3Si2 particles to an atmosphere of up to 15% oxygen by volume dispersed in an inert gas for a period of time and at a temperature sufficient to form UO2 at the U3Si2 particle surface, or by mixing U3Si2 particles with an amount up to 30% by weight of UO2 particles.