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Defect cluster and nonequilibrium gas bubble associated growth in irradiated UMo fuels – A cluster dynamics and phase field model

Journal of nuclear materials, 2020-12, Vol.542, p.152441, Article 152441

2020

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A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

Journal of nuclear materials, 2009-06, Vol.389 (3), p.458-469

2009

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Thermo-physical properties of DU–10 wt.% Mo alloys

Journal of nuclear materials, 2010-08, Vol.403 (1), p.160-166

2010

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Metallic fast reactor fuel fabrication for the global nuclear energy partnership

Journal of nuclear materials, 2009-07, Vol.392 (2), p.158-163

2009

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Modeling thermal and stress behavior of the fuel–clad interface in monolithic fuel mini-plates

Materials in engineering, 2010-08, Vol.31 (7), p.3234-3243

2010

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Determination of the degree of grain refinement in irradiated U-Mo fuels

Heliyon, 2018-12, Vol.4 (12), p.e00920-e00920, Article e00920

2018

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Mechanical Properties of DU-xMo Alloys with x = 7 to 12 Weight Percent

Metallurgical and materials transactions. A, Physical metallurgy and materials science, 2009-05, Vol.40 (5), p.1069-1079

2009

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A Monte Carlo model of irradiation-induced recrystallization in polycrystalline UMo fuels

Journal of nuclear materials, 2019-10, Vol.524 (C), p.164-176

2019

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Densification of nano-CeO2 ceramics as nuclear oxide surrogate by spark plasma sintering

Journal of nuclear materials, 2010-09, Vol.404 (3), p.210-216

2010

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Assessment of effective thermal conductivity in U–Mo metallic fuels with distributed gas bubbles

Journal of nuclear materials, 2015-07, Vol.462 (C), p.64-76

2015

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Finite-element model to predict roll-separation force and defects during rolling of U-10Mo alloys

Journal of nuclear materials, 2017-10, Vol.494 (C), p.182-191

2017

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Modeling the influence of interaction layer formation on thermal conductivity of U–Mo dispersion fuel

Journal of alloys and compounds, 2015-01, Vol.618, p.7-13

2015

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Thermomechanical process optimization of U-10wt% Mo – Part 2: The effect of homogenization on the mechanical properties and microstructure

Journal of nuclear materials, 2015-10, Vol.465 (C), p.710-718

2015

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Thermal properties of U-Mo alloys irradiated under high fission power density

Journal of nuclear materials, 2021-04, Vol.547, p.152823, Article 152823

2021

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Phase characteristics of a number of U–Pu–Am–Np–Zr metallic alloys for use as fast reactor fuels

Journal of nuclear materials, 2010, Vol.396 (1), p.49-56

2010

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The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

Journal of nuclear materials, 2017-04, Vol.486, p.222-233

2017

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Thermal properties of U–Mo alloys irradiated to moderate burnup and power

Journal of nuclear materials, 2015-09, Vol.464 (C), p.331-341

2015

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A model to predict failure of irradiated U–Mo dispersion fuel

Nuclear engineering and design, 2016-12, Vol.310 (C), p.48-56

2016

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Thermal conductivity of fresh and irradiated U-Mo fuels

Journal of nuclear materials, 2018-05, Vol.503 (C), p.304-313

2018

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Gray Phonon Transport Prediction of Thermal Conductivity in Lithium Aluminate with Higher-Order Finite Elements on Meshes with Curved Surfaces

Journal of computational and theoretical transport, 2021-07, Vol.50 (5), p.483-506

2021

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Microstructural characteristics of DU– xMo alloys with x = 7–12 wt

Journal of alloys and compounds, 2009-06, Vol.479 (1), p.140-147

2009

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Image fusion of Secondary Ion Mass Spectrometry and Energy-dispersive X-Ray Spectroscopy data for the characterization of uranium-molybdenum fuel foils

Journal of nuclear materials, 2018-01, Vol.498 (C), p.348-354

2018

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Thermo-mechanical behavior simulation coupled with the hydrostatic-pressure-dependent grain-scale fission gas swelling calculation for a monolithic UMo fuel plate under heterogeneous neutron irradiation

Open Engineering (Warsaw), 2018-08, Vol.8 (1), p.243-260

2018

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Properties of DU–10 wt% Mo alloys subjected to various post-rolling heat treatments

Nuclear engineering and design, 2010-06, Vol.240 (6), p.1332-1339

2010

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Measurement of fission gas release from irradiated U–Mo monolithic fuel samples

Journal of nuclear materials, 2015-06, Vol.461, p.61-71

2015

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Defect cluster and nonequilibrium gas bubble associated growth in irradiated UMo fuels – A cluster dynamics and phase field model

A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

Thermo-physical properties of DU–10 wt.% Mo alloys

Metallic fast reactor fuel fabrication for the global nuclear energy partnership

Modeling thermal and stress behavior of the fuel–clad interface in monolithic fuel mini-plates

Determination of the degree of grain refinement in irradiated U-Mo fuels

Mechanical Properties of DU-xMo Alloys with x = 7 to 12 Weight Percent

A Monte Carlo model of irradiation-induced recrystallization in polycrystalline UMo fuels

Densification of nano-CeO2 ceramics as nuclear oxide surrogate by spark plasma sintering

Assessment of effective thermal conductivity in U–Mo metallic fuels with distributed gas bubbles

Finite-element model to predict roll-separation force and defects during rolling of U-10Mo alloys

Modeling the influence of interaction layer formation on thermal conductivity of U–Mo dispersion fuel

Thermomechanical process optimization of U-10wt% Mo – Part 2: The effect of homogenization on the mechanical properties and microstructure

Thermal properties of U-Mo alloys irradiated under high fission power density

Phase characteristics of a number of U–Pu–Am–Np–Zr metallic alloys for use as fast reactor fuels

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

Thermal properties of U–Mo alloys irradiated to moderate burnup and power

A model to predict failure of irradiated U–Mo dispersion fuel

Thermal conductivity of fresh and irradiated U-Mo fuels

Gray Phonon Transport Prediction of Thermal Conductivity in Lithium Aluminate with Higher-Order Finite Elements on Meshes with Curved Surfaces

Microstructural characteristics of DU– xMo alloys with x = 7–12 wt

Image fusion of Secondary Ion Mass Spectrometry and Energy-dispersive X-Ray Spectroscopy data for the characterization of uranium-molybdenum fuel foils

Thermo-mechanical behavior simulation coupled with the hydrostatic-pressure-dependent grain-scale fission gas swelling calculation for a monolithic UMo fuel plate under heterogeneous neutron irradiation

Properties of DU–10 wt% Mo alloys subjected to various post-rolling heat treatments

Measurement of fission gas release from irradiated U–Mo monolithic fuel samples

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