Westinghouse Electric Company LLC

Kabushiki Kaisha Toshiba Aka Toshiba Corporation

Wisconsin Alumni Research Foundation

Toshiba Energy Systems & Solutions Corporation

Westinghouse Electric Sweden Ab

Peng Xu

Edward Jean Lahoda

Lu Cai

Jason P Mazzoccoli

Robert L Oelrich

Kumar Sridharan

Benjamin R Maier

Greg Johnson

Sumit Ray

Lars Hallstadius

Zeses E Karoutas

Hemant H Shah

Fumihisa Kano

Jonathan Wright

Joon Hyung Choi

Yun Long

John L Lyons

Carroll J Long, Jr

Paolo Ferroni

Frank A Boylan

Michael R Ickes

Shinichi Higuchi

Fausto Franceschini

Zeses Karoutas

Hwasung Yeom

Grant L Eddy

Shinichi Higuichi

Hemant Shah

Benjamin Maier

Javier Romero

Kenneth GÖRANSSON

Summit Ray

An improvement in a nuclear fuel rod is disclosed. The improved fuel rod includes a cladding tube, a plurality of fuel pellets stacked within the cladding tube, and a liquid material filling the gap between the fuel pellets and the cladding tube. The liquid material is selected from those having a thermal conductivity higher than that of helium, a melting point lower than about 400° C., a boiling point higher than 1600° C., and which are capable of wetting both the fuel pellets and the cladding sufficient to form a protective layer over the pellets and to wick into openings that may form in the cladding.

Fuel rod having heat transfer liquid filling the gap between fuel pellets and cladding tube

An elongate fuel element is described that has a silicon carbide cladding enclosing a fuel, such as UO, wherein the fuel is dimensioned relative to the cladding to define gaps at each lateral end of the enclosure sufficiently large such that upon swelling in use, the fuel does not increase the strain on the cladding beyond the limits of the claddings strain tolerance. The lateral gaps at the ends of the fuel allow lateral expansion during swelling that reduces the strain on the cladding.

Elongate SiC fuel elements

Disclosed is a method for manufacturing free-standing cladding tubes with multi-layer structures. According to the method, a cylindrical mandrel substrate defining a hollow cylindrical inner space is provided. A first cold spray powder metal is selected. The cylindrical mandrel substrate is rotated and the first cold spray powder metal is applied to an outer surface of the cylindrical mandrel substrate to form a first layer. The cylindrical mandrel substrate is removed.

Methods for manufacturing nanostructured and compositionally-tailored tubes and components by low temperature, solid-state cold spray powder deposition

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 USiwith 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 USi, then the oxidation resistant particles can have a particle size distribution less than that of the UN or USi.

Grain boundary enhanced UN and USipellets with improved oxidation resistance

Disclosed herein is a method comprising coating a fissile, uranium-containing ceramic material with a water-resistant layer, the layer being non-reactive with the fissile, uranium-containing ceramic material. The coating is applied to a surface of the fissile, uranium-containing ceramic material. Also disclosed is a fuel for use in a nuclear reactor.

Coated fuel pellets with enhanced water and steam oxidation resistance

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.

Cold spray chromium coating for nuclear fuel rods

Grain boundary enhanced UN and U3Si2 pellets with improved oxidation resistance

An apparatus and method for pressurizing SiC clad rods of a nuclear core component. A lower end of the rod is sealed with a lower end plug and an upper end of the rod is sealed between the cladding and an external piece of an upper end plug that has a through opening through which a separate internal piece of the upper end plug extends. The internal piece of the upper end plug is initially moveable within the through opening between an upper position that forms a gas tight seal and a lower position that forms a gaseous path through the through opening. The rod is placed in a pressure chamber pressurized to a desired pressure. When the pressure is reduced within the pressure chamber the internal pressure in the rod biases the internal piece of the upper end plug in the upper sealed position.

Method to pressurize sic fuel cladding tube before end plug sealing by pressurization pushing spring loaded end plug

A method is described for forming duplex layers including an interlayer and a corrosion resistant boundary layer on a nuclear fuel rod cladding tube for use in a water cooled nuclear reactor. The method includes forming, by thermal deposition or physical vapor deposition, on the exterior of a substrate, an inner interlayer with Mo, Ta, W or Nb or other particles, and forming, by thermal deposition or physical vapor deposition, on the interlayer, an outer corrosion resistant layer with particles selected from the group consisting of Cr, a Cr alloy, and combinations thereof. The interlayer prevents eutectic formation between the corrosion resistant layer and the substrate.

Duplex accident tolerant coating for nuclear fuel rods

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.

Deposition of integrated protective material into zirconium cladding for nuclear reactors by high-velocity thermal application

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 coating the fissile material, such as a pellet of USiand/or the grain boundaries, to a desired thickness with a suitable coating material, such as atomic layer deposition or a thermal spray process. The coating material may be any non-reactive material with a solubility at least as low as that of UO. Exemplary coating materials include ZrSiO, FeCrAl, Cr, Zr, Al&#x2014;Cr, CrAl, ZrO, CeO, TiO, SiO, UO, ZrB, NaO&#x2014;BO&#x2014;SiO&#x2014;AlOglass, AlO, CrO, carbon, and SiC, and combinations thereof. The water resistant layer may be overlayed with a burnable absorber layer, such as ZrBor BO&#x2014;SiOglass.

Coated U3Si2 pellets with enhanced water and steam oxidation resistance

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&#xb0; C. and 1200&#xb0; 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.

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&#xb0; C. to 600&#xb0; 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.

Method of manufacturing a SiC composite fuel cladding with inner Zr alloy liner

The invention relates to compositions and methods for coating a zirconium alloy cladding of a fuel element for a nuclear water reactor. The composition includes a master alloy including one or more alloying elements selected from chromium, silicon and aluminum, a chemical activator and an inert filler. The alloying element(s) is deposited or are co-deposited on the cladding using a pack cementation process. When the coated zirconium alloy cladding is exposed to and contacted with water in a nuclear reactor, a protective oxide layer can form on the coated surface of the cladding.

Corrosion and wear resistant coating on zirconium alloy cladding

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.

Growing community of inventors

Columbia, SC, United States of America

Peng Xu