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High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments

Journal of nuclear materials, 2013-09, Vol.440 (1-3), p.420-427

2013

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Thermo-mechanical analysis of LWR SiC/SiC composite cladding

Journal of nuclear materials, 2014-04, Vol.447 (1-3), p.125-142

2014

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Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

Journal of nuclear materials, 2014-05, Vol.448 (1-3), p.420-435

2014

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Fabrication and characterization of fully ceramic microencapsulated fuels

Journal of nuclear materials, 2012-07, Vol.426 (1-3), p.268-276

2012

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Effect of steam on high temperature oxidation behaviour of alumina-forming alloys

Materials at high temperatures, 2015-01, Vol.32 (1-2), p.28-35

2015

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Accident tolerant fuels for LWRs: A perspective

Journal of nuclear materials, 2014-05, Vol.448 (1-3), p.374-379

2014

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Evaluation of the Mechanical Properties of TRISO Particles Using Nanoindentation and Ring Compression Testing

Experimental mechanics, 2017-09, Vol.57 (7), p.1081-1090

2017

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Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

Journal of nuclear materials, 2015-12, Vol.467 (Part 2), p.703-716

2015

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The kinetics of hydrogen desorption from and adsorption on zirconium hydride

Journal of nuclear materials, 2010-02, Vol.397 (1), p.61-68

2010

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Application of X‐ray microcomputed tomography in the characterization of irradiated nuclear fuel and material specimens

Journal of microscopy (Oxford), 2015-11, Vol.260 (2), p.163-174

2015

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An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

Journal of nuclear materials, 2017-08, Vol.491 (C), p.55-66

2017

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In situ ceramic layer growth on coated fuel particles dispersed in a zirconium metal matrix

Journal of nuclear materials, 2013-06, Vol.437 (1-3), p.171-177

2013

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Irradiation-enhanced α′ precipitation in model FeCrAl alloys

Scripta materialia, 2016-04, Vol.116 (C), p.112-116

2016

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Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form

Journal of nuclear materials, 2015-02, Vol.457, p.9-17

2015

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Effect of post weld heat treatment on the 6061 aluminum alloy produced by ultrasonic additive manufacturing

Materials science & engineering. A, Structural materials : properties, microstructure and processing, 2017-01, Vol.684 (C), p.606-616

2017

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Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions

Journal of nuclear materials, 2014-05, Vol.448 (1-3), p.389-398

2014

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Irradiation stability and thermomechanical properties of 3D-printed SiC

Journal of nuclear materials, 2021-08, Vol.551 (C), p.152980, Article 152980

2021

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Architecture and properties of TCR fuel form

Journal of nuclear materials, 2021-04, Vol.547 (-), p.152781, Article 152781

2021

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Modeling reactivity insertion experiments of TRISO particles in NSRR using BISON

Journal of nuclear materials, 2020-03, Vol.530 (C), p.151965, Article 151965

2020

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Transformational Challenge Reactor preconceptual core design studies

Nuclear engineering and design, 2020-10, Vol.367, p.110781, Article 110781

2020

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Influence of neutron irradiation on Al-6061 alloy produced via ultrasonic additive manufacturing

Journal of nuclear materials, 2021-07, Vol.550 (C), p.152939, Article 152939

2021

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Fabrication of UO2-Mo composite fuel with enhanced thermal conductivity from sol-gel feedstock

Journal of nuclear materials, 2019-07, Vol.520 (C), p.56-64

2019

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Performance of U3Si2 in an LWR following a cladding breach during normal operation

Journal of nuclear materials, 2020-10, Vol.539 (C), p.152263, Article 152263

2020

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Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

Journal of nuclear materials, 2021-05, Vol.548, p.152849, Article 152849

2021

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Scoping assessments of ATF impact on late-stage accident progression including molten core–concrete interaction

Journal of nuclear materials, 2014-05, Vol.448 (1-3), p.534-540

2014

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High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments

Thermo-mechanical analysis of LWR SiC/SiC composite cladding

Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

Fabrication and characterization of fully ceramic microencapsulated fuels

Effect of steam on high temperature oxidation behaviour of alumina-forming alloys

Accident tolerant fuels for LWRs: A perspective

Evaluation of the Mechanical Properties of TRISO Particles Using Nanoindentation and Ring Compression Testing

Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

The kinetics of hydrogen desorption from and adsorption on zirconium hydride

Application of X‐ray microcomputed tomography in the characterization of irradiated nuclear fuel and material specimens

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

In situ ceramic layer growth on coated fuel particles dispersed in a zirconium metal matrix

Irradiation-enhanced α′ precipitation in model FeCrAl alloys

Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form

Effect of post weld heat treatment on the 6061 aluminum alloy produced by ultrasonic additive manufacturing

Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions

Irradiation stability and thermomechanical properties of 3D-printed SiC

Architecture and properties of TCR fuel form

Modeling reactivity insertion experiments of TRISO particles in NSRR using BISON

Transformational Challenge Reactor preconceptual core design studies

Influence of neutron irradiation on Al-6061 alloy produced via ultrasonic additive manufacturing

Fabrication of UO2-Mo composite fuel with enhanced thermal conductivity from sol-gel feedstock

Performance of U3Si2 in an LWR following a cladding breach during normal operation

Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

Scoping assessments of ATF impact on late-stage accident progression including molten core–concrete interaction

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