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Quasi-Diffusion Method with 3-D Cross Sections for TREAT Core Analysis

Nuclear technology, 2020-06, Vol.206 (6), p.825-838

2020

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Development and Analysis of TREAT Neutronics Benchmarks

Nuclear technology, 2020-06, Vol.206 (6), p.805-824

2020

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Multi-group SP3 approximation for simulation of a three-dimensional PWR rod ejection accident

Annals of nuclear energy, 2015-03, Vol.77, p.94-100

2015

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Surrogate-based multi-experiment calibration of the BISON fission gas behavior model

Nuclear engineering and design, 2017-08, Vol.320, p.409-417

2017

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Impact of improved neutronic methodology on the cladding response during a PWR reactivity initiated accident

Nuclear engineering and design, 2013-09, Vol.262, p.180-188

2013

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SP3 Limit of the 2D/1D Transport Equations with Varying Degrees of Angular Coupling

Journal of computational and theoretical transport, 2020-09, Vol.49 (6), p.303-330

2020

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Cell Homogenization Method for Pin-by-Pin Neutron Transport Calculations

Nuclear science and engineering, 2011-09, Vol.169 (1), p.1-18

2011

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Assessment of assembly homogenized two-steps core dynamic calculations using direct whole core transport solutions

Annals of nuclear energy, 2016-01, Vol.87, p.356-365

2016

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Validation of the U.S. NRC Coupled Code System TRITON/TRACE/PARCS Using the Special Power Excursion Reactor Test III

Nuclear technology, 2013-09, Vol.183 (3), p.504-514

2013

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The treatment of resonance interference effects in the subgroup method

Annals of nuclear energy, 2011-05, Vol.38 (5), p.995-1003

2011

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An endothermic chemical process facility coupled to a high temperature reactor. Part I: Proposed accident scenarios within the chemical plant

Nuclear engineering and design, 2012-05, Vol.246, p.256-265

2012

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Methods and Model Development for Coupled RELAP5/PARCS Analysis of the Atucha-II Nuclear Power Plant

Science and Technology of Nuclear Installations, 2011-01, Vol.2011 (1)

2011

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Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

Nuclear engineering and design, 2012-07, Vol.248, p.1-13

2012

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Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

Nuclear technology, 2002-04, Vol.138 (1), p.17-29

2002

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Multi-level coarse mesh finite difference acceleration with local two-node nodal expansion method

Annals of nuclear energy, 2018-06, Vol.116, p.105-113

2018

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Implementation of Two-Level Coarse-Mesh Finite Difference Acceleration in an Arbitrary Geometry, Two-Dimensional Discrete Ordinates Transport Method

Nuclear science and engineering, 2008-03, Vol.158 (3), p.289-298

2008

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High-Fidelity Light Water Reactor Analysis with the Numerical Nuclear Reactor

Nuclear science and engineering, 2007-03, Vol.155 (3), p.395-408

2007

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Letter to the Editor on the paper by W. Wulff, “Critical review of conservation equations for two-phase flow in the U.S. NRC TRACE Code”

Nuclear engineering and design, 2011-10, Vol.241 (10), p.4261-4268

2011

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Parameter Sensitivity Study of Boiling and Two-Phase Flow Models in CFD

The journal of computational multiphase flows (Online), 2012-12, Vol.4 (4), p.411-425

2012

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The Legendre Polynomial Axial Expansion Method

Nuclear science and engineering, 2023-02, Vol.197 (2), p.291-307

2023

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VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

Nuclear Engineering and Technology, 2013, 45(6), , pp.745-752

2013

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The Legendre Polynomial Axial Expansion Method

Nuclear science and engineering, 2022-07, Vol.197 (2)

2022

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Convergence Analysis of the Nonlinear Coarse-Mesh Finite Difference Method for One-Dimensional Fixed-Source Neutron Diffusion Problem

Nuclear science and engineering, 2004-06, Vol.147 (2), p.127-147

2004

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The Subray Method of Characteristics

Nuclear science and engineering, 2019-06, Vol.193 (6), p.601-621

2019

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Error Assessment of Homogenized Cross Sections Generation for Whole Core Neutronic Calculation

AIP conference proceedings, 2007, Vol.947 (1), p.39-42

2007

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Quasi-Diffusion Method with 3-D Cross Sections for TREAT Core Analysis

Development and Analysis of TREAT Neutronics Benchmarks

Multi-group SP3 approximation for simulation of a three-dimensional PWR rod ejection accident

Surrogate-based multi-experiment calibration of the BISON fission gas behavior model

Impact of improved neutronic methodology on the cladding response during a PWR reactivity initiated accident

SP3 Limit of the 2D/1D Transport Equations with Varying Degrees of Angular Coupling

Cell Homogenization Method for Pin-by-Pin Neutron Transport Calculations

Assessment of assembly homogenized two-steps core dynamic calculations using direct whole core transport solutions

Validation of the U.S. NRC Coupled Code System TRITON/TRACE/PARCS Using the Special Power Excursion Reactor Test III

The treatment of resonance interference effects in the subgroup method

An endothermic chemical process facility coupled to a high temperature reactor. Part I: Proposed accident scenarios within the chemical plant

Methods and Model Development for Coupled RELAP5/PARCS Analysis of the Atucha-II Nuclear Power Plant

Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

Multi-level coarse mesh finite difference acceleration with local two-node nodal expansion method

Implementation of Two-Level Coarse-Mesh Finite Difference Acceleration in an Arbitrary Geometry, Two-Dimensional Discrete Ordinates Transport Method

High-Fidelity Light Water Reactor Analysis with the Numerical Nuclear Reactor

Letter to the Editor on the paper by W. Wulff, “Critical review of conservation equations for two-phase flow in the U.S. NRC TRACE Code”

Parameter Sensitivity Study of Boiling and Two-Phase Flow Models in CFD

The Legendre Polynomial Axial Expansion Method

VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

The Legendre Polynomial Axial Expansion Method

Convergence Analysis of the Nonlinear Coarse-Mesh Finite Difference Method for One-Dimensional Fixed-Source Neutron Diffusion Problem

The Subray Method of Characteristics

Error Assessment of Homogenized Cross Sections Generation for Whole Core Neutronic Calculation

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