Westinghouse Electric Company LLC

Westinghouse Electric Corporation

The United States of America as Represented by the United States

Westinghouse Electric Sweden Ab

Kabushiki Kaisha Toshiba Aka Toshiba Corporation

Wisconsin Alumni Research Foundation

Toshiba Energy Systems & Solutions Corporation

Edward Jean Lahoda

Peng Xu

David C Grant

Sumit Ray

Thomas V Congedo

Jason P Mazzoccoli

Robert L Oelrich

Thomas S Snyder

Lars Hallstadius

Steven H Peterson

Radu O Pomirleanu

John Bartko

Zeses E Karoutas

Michael C Skriba

Kumar Sridharan

Warren R Junker

Lyman J Petrosky

Lu Cai

Edward F Sverdrup

Herbert A Burgman

Benjamin R Maier

Greg Johnson

John P Lareau

Ching-Yu Lin

Ernest D Lee

Michael J Hone

James M Pope

Paolo Ferroni

Hemant H Shah

Robert E Witkowski

William A Byers

Deborah P Partlow

Michael Anthony Burke

David F McLaughlin

John G Seidel

William M Alvino

Robert D Burack

Yasir Arafat

Young J Kwon

Mario D Carelli

Frank H Ruddy

Mats Dahlbäck

Arthur L Wolfe

Frank A Boylan

David A Eckhardt

Ho Q Lam

Jonathan Wright

Albert J Dietrich

Sharon L Weisberg

Joon Hyung Choi

Fumihisa Kano

William D Partlow

Leland L Learn

Mitchell E Nissley

Yun Long

Raymond E Schneider

Fausto Franceschini

Thomas H Dent

Jurie Van Wyk

Peter T Chiang

Ivan A Martorell

John L Lyons

Beryl H Parks

Matthew J Memmott

Simon Charles Middleburgh

Don E Harrison

Joseph L Gonzalez

Andrew J Mueller

Robert J Leduc

Robert W White

Shinichi Higuchi

Arthur W Ott

Donald L Grover

Zeses Karoutas

Kathryn E Metzger

Hans Widegren

Francis Talko

Clinton B Armstrong

Nicholas D'Amico

Donald W Cash

Jerry C Seagle

C Keith Paulson

John A Paffenbarger

A J Impink, Jr

J A Battaglia

C Y Lin

C P Keegan

David F McLaughin

Lars Hallstadias

Keith D Task

Hwasung Yeom

Luke C Olson

Hemant Shah

Khoa T Ha

Shinichi Higuichi

Benjamin Maier

Mikael Jolkkonen

Javier Romero

Kenneth GÖRANSSON

Nuclear fuel cladding for fast reactors, assemblies thereof, and methods of manufacture thereof are provided. The nuclear fuel cladding comprises a substrate, a first layer, and a second layer. The substrate comprises a tubular shape. The first layer is deposited over an external surface of the substrate. The first layer comprises a corrosion resistant composition. The second layer is disposed over the first layer. The second layer comprises silicon carbide fibers infiltrated with silicon carbide. The second layer is configured to inhibit outward creep of the substrate. A gap is defined intermediate the first layer and the second layer.

Nuclear fuel cladding for fast reactors

A covering for reinforcing a base layer of a nuclear fuel cladding is provided. The covering comprises a first layer configured to cover a first portion of the outer surface of the base layer of the nuclear fuel cladding, a second layer surrounding the first layer and the base layer and a third layer surrounding the second layer. The first layer comprises a fiber based material, the second layer comprises an interfacing material and the third layer comprises Chromium. A reinforced cladding for nuclear fuel and a method for producing a reinforced nuclear fuel cladding are also provided.

Fiber reinforced multi-layered wear and corrosion coatings of zirconium alloy nuclear fuel cladding

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

The present application relates to nuclear fuel claddings, and more particularly to zirconium alloy tubes with a ceramic covering. A first coating layer is applied to the ceramic covering. A second coating layer is over the first coating layer. The first coating layer includes Nb. The second coating layer includes Cr.

Silicon carbide reinforced zirconium based cladding

A control rod for a nuclear fuel assembly is described herein that includes a neutron absorbing material having a melting point greater than 1500° C. that does not form a eutectic with a melting point less than 1500° C., and may further include a cladding material having a melting point greater than 1500° C. The cladding material is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material is selected from the group consisting of GdO, Ir, BC, Re, and Hf. The metal cladding or the absorbing material may be coated with an anti-oxidation coating of Cr with or without a Nb intermediate layer.

Control rods for light water reactors

A method is described herein that produces UN from UFin at most two steps comprising UF→intermediate→UN. The principle of the reaction is that in a first step, UFwould be reduced to UN, where x may be an integer selected from 1 and 3, and y is an integer selected from 1 and 2. Reduction occurs at or near the surface of a gaseous membrane electrode where it is also in contact with a nitrogen bearing salt. In a second step, UNdecomposes to UN and Ngas, either in the same reactor as the first step or after removal to a separate unit for further processing.

Electrochemical uranium nitride production

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

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

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&#xb0; C.-1000&#xb0; 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.

A control rod for a nuclear fuel assembly is described herein that includes a neutron absorbing material having a melting point greater than 1500&#xb0; C. that does not form a eutectic with a melting point less than 1500&#xb0; C., and may further include a cladding material having a melting point greater than 1500&#xb0; C. The cladding material is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material is selected from the group consisting of GdO, Ir, BC, Re, and Hf. The metal cladding or the absorbing material may be coated with an anti-oxidation coating of Cr with or without a Nb intermediate layer.

Grain boundary enhanced UN and U3Si2 pellets with improved oxidation resistance

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.

Manufacture of large grain powders with granular coatings

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

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Edgewood, PA, United States of America

Edward Jean Lahoda