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Journal Articles

Pearson Correlation and Chi-Squared Methods Applied to the Sensitivity and Uncertainty Analysis of Hydrogen Generation During a BWR STSBO Using MAAP5 and AZTUSIA

Javier Ortiz-Villafuerte, Heriberto Sanchez-Mora, Cesar Queral, Melisa Reyes-Fuentes, Edmundo del-Valle-Gallegos

ASME J of Nuclear Rad Sci.

Paper No: NERS-22-1021

DOI: https://doi.org/10.1115/1.4055011

Published Online: July 15, 2022

View Article titled, Pearson Correlation and Chi-Squared Methods Applied to the Sensitivity and Uncertainty Analysis of Hydrogen Generation During a BWR STSBO Using MAAP5 and AZTUSIA

Topics: Boiling water reactors, Hydrogen, Uncertainty analysis, Accidents, Simulation, Explosions, Cladding systems (Building), Failure, Fukushima nuclear disaster, Japan, 2011, Nuclear power stations

Journal Articles

Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

Emmanuel K. Boafo, Emmanuel Numapau Gyamfi

ASME J of Nuclear Rad Sci. July 2022, 8(3): 031703.

Paper No: NERS-21-1094

DOI: https://doi.org/10.1115/1.4053050

Published Online: May 26, 2022

View Article titled, Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

Topics: Accidents, Hydrogen, Sensitivity analysis, Uncertainty, Uncertainty quantification, Uncertainty analysis, Design, Optimization

Journal Articles

The Impact of Fueling Operations on Full Core Uncertainty Analysis in CANDU Reactors

M. D. Tucker, D. R. Novog

ASME J of Nuclear Rad Sci. July 2020, 6(3): 031401.

Paper No: NERS-19-1065

DOI: https://doi.org/10.1115/1.4045485

Published Online: March 9, 2020

View Article titled, The Impact of Fueling Operations on Full Core Uncertainty Analysis in CANDU Reactors

Topics: Algorithms, Cross section (Physics), Feedback, Fuels, Simulation, Uncertainty, Uncertainty analysis, Uranium, Physics, Diffusion (Physics)

Journal Articles

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

Massimo Sarotto, Gabriele Firpo, Anatoly Kochetkov, Antonin Krása, Emil Fridman, Jerzy Cetnar, Grazyna Domanska

ASME J of Nuclear Rad Sci. January 2020, 6(1): 011402.

Paper No: NERS-18-1132

DOI: https://doi.org/10.1115/1.4044000

Published Online: November 29, 2019

View Article titled, Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

Topics: Design, Fast neutron reactors, Nuclear fission, Nuclides, Uncertainty, Uncertainty analysis, Licensing, Neutrons, Rods, Simulation

Journal Articles

Development of a Seismic Probabilistic Safety Assessment Quantification Tool

Wei Gao, Guofeng Tang, Jingyu Zhang, Qinfang Zhang

ASME J of Nuclear Rad Sci. October 2018, 4(4): 041005.

Paper No: NERS-17-1148

DOI: https://doi.org/10.1115/1.4039967

Published Online: September 10, 2018

View Article titled, Development of a Seismic Probabilistic Safety Assessment Quantification Tool

Topics: Algorithms, Computer software, Earthquake risk, Nuclear engineering, Probabilistic safety assessment, Probability, Uncertainty analysis, Nuclear power stations, China, Fukushima nuclear disaster, Japan, 2011

Journal Articles

Uncertainty Analysis for Source Term Evaluation of High Temperature Gas-Cooled Reactor Under Accident Conditions—Identification of Influencing Factors in Loss-of-Forced Circulation Accidents

Yuki Honda, Hiroyuki Sato, Shigeaki Nakagawa, Hirofumi Ohashi

ASME J of Nuclear Rad Sci. July 2018, 4(3): 031013.

Paper No: NERS-17-1253

DOI: https://doi.org/10.1115/1.4039066

Published Online: May 16, 2018

View Article titled, Uncertainty Analysis for Source Term Evaluation of High Temperature Gas-Cooled Reactor Under Accident Conditions—Identification of Influencing Factors in Loss-of-Forced Circulation Accidents

Topics: Accidents, Fuels, Temperature, Uncertainty, Uncertainty analysis, Very high temperature reactors, Heat, Failure

Journal Articles

Method and Validation for Measurement of Effective Thermal Diffusivity and Conductivity of Pebble Bed in High Temperature Gas-Cooled Reactors

Yongyong Wu, Cheng Ren, Rui Li, Xingtuan Yang, Jiyuan Tu, Shengyao Jiang

ASME J of Nuclear Rad Sci. July 2018, 4(3): 031006.

Paper No: NERS-17-1109

DOI: https://doi.org/10.1115/1.4039035

Published Online: May 16, 2018

View Article titled, Method and Validation for Measurement of Effective Thermal Diffusivity and Conductivity of Pebble Bed in High Temperature Gas-Cooled Reactors

Topics: Errors, Temperature, Thermal conductivity, Thermal diffusivity, Very high temperature reactors, Low temperature, Safety, Heat conduction, Algorithms, Electrical conductivity

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Pearson Correlation and Chi-Squared Methods Applied to the Sensitivity and Uncertainty Analysis of Hydrogen Generation During a BWR STSBO Using MAAP5 and AZTUSIA

Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

The Impact of Fueling Operations on Full Core Uncertainty Analysis in CANDU Reactors

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

Development of a Seismic Probabilistic Safety Assessment Quantification Tool

Uncertainty Analysis for Source Term Evaluation of High Temperature Gas-Cooled Reactor Under Accident Conditions—Identification of Influencing Factors in Loss-of-Forced Circulation Accidents

Method and Validation for Measurement of Effective Thermal Diffusivity and Conductivity of Pebble Bed in High Temperature Gas-Cooled Reactors

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Pearson Correlation and Chi-Squared Methods Applied to the Sensitivity and Uncertainty Analysis of Hydrogen Generation During a BWR STSBO Using MAAP5 and AZTUSIA

Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

The Impact of Fueling Operations on Full Core Uncertainty Analysis in CANDU Reactors

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

Development of a Seismic Probabilistic Safety Assessment Quantification Tool

Uncertainty Analysis for Source Term Evaluation of High Temperature Gas-Cooled Reactor Under Accident Conditions—Identification of Influencing Factors in Loss-of-Forced Circulation Accidents

Method and Validation for Measurement of Effective Thermal Diffusivity and Conductivity of Pebble Bed in High Temperature Gas-Cooled Reactors

Journals

Conference Proceedings

eBooks

Resources

Opportunities

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