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Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

Journal of Nuclear Engineering, 2021-09, Vol.2 (4), p.318-335

2021

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DEVELOPMENT OF MPACT FOR FULL-CORE SIMULATIONS OF MAGNOX GAS-COOLED NUCLEAR REACTORS

EPJ Web of conferences, 2021-01, Vol.247, p.6041

2021

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A Full-Core Resonance Self-Shielding Method Using a Continuous-Energy Quasi-One-Dimensional Slowing-Down Solution That Accounts for Temperature-Dependent Fuel Subregions and Resonance Interference

Nuclear science and engineering, 2015-07, Vol.180 (3), p.247-272

2015

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Whole Core Transport Calculation Employing Hexagonal Modular Ray Tracing and CMFD Formulation

Journal of nuclear science and technology, 2008-08, Vol.45 (8), p.740-751

2008

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Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor

Journal of Nuclear Engineering, 2024-08, Vol.5 (3), p.260-273

2024

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Iterative resonance self-shielding methods using resonance integral table in heterogeneous transport lattice calculations

Annals of nuclear energy, 2011, Vol.38 (1), p.32-43

2011

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Validation of the SCALE/Polaris−PARCS Code Procedure with the ENDF/B-VII.1 AMPX 56-Group Library: Pressurized Water Reactor

Journal of Nuclear Engineering, 2024-07, Vol.5 (3), p.246-259

2024

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DEVELOPMENT OF THE MPACT 69-GROUP LIBRARY FOR MAGNOX REACTOR ANALYSIS USING CASL VERA

EPJ Web of conferences, 2021-01, Vol.247, p.10016

2021

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MPACT VERIFICATION WITH MAGNOX REACTOR NEUTRONICS PROGRESSION PROBLEMS

EPJ Web of conferences, 2021-01, Vol.247, p.10031

2021

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IMPROVEMENT OF SCALE-XSProc MULTIGROUP CROSS SECTION PROCESSING BASED ON THE CENTRM POINTWISE SLOWING DOWN CALCULATION

EPJ Web of conferences, 2021, Vol.247, p.2011

2021

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AN IMPROVED ENERGY DEPOSITION MODEL IN MPACT AND EXPLICIT HEAT GENERATION COUPLING WITH CTF

EPJ Web of conferences, 2021-01, Vol.247, p.2033

2021

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Gamma transport and diffusion calculation capability coupled with neutron transport simulation in KARMA 1.2

Annals of nuclear energy, 2013-07, Vol.57, p.59-67

2013

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A new procedure to generate resonance integral table with an explicit resonance interference for transport lattice codes

Annals of nuclear energy, 2011, Vol.38 (1), p.118-127

2011

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Unstructured partial- and net-current based coarse mesh finite difference acceleration applied to the extended step characteristics method in NEWT

Annals of nuclear energy, 2011-02, Vol.38 (2), p.527-534

2011

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The method of characteristics applied to solving slowing down equation to estimate the self-shielded resonance cross sections with an explicit geometrical effect

Annals of nuclear energy, 2011-02, Vol.38 (2), p.438-446

2011

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Fourier Convergence Analysis of the Rebalance Methods for Discrete Ordinates Transport Equations in Eigenvalue Problems

Nuclear science and engineering, 2010-01, Vol.164 (1), p.33-52

2010

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Axial SPN and Radial MOC Coupled Whole Core Transport Calculation

Journal of nuclear science and technology, 2007-09, Vol.44 (9), p.1156-1171

2007

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Development of a computer code system for the analysis of prism and pebble type VHTR cores

Annals of nuclear energy, 2008-10, Vol.35 (10), p.1919-1928

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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Development of a physics analysis procedure for the prismatic very high temperature gas-cooled reactors

Annals of nuclear energy, 2007-11, Vol.34 (11), p.849-860

2007

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IAEA GT-MHR benchmark calculations by using the HELIOS/MASTER physics analysis procedure and the MCNP Monte Carlo code

Nuclear engineering and design, 2008-10, Vol.238 (10), p.2654-2667

2008

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Verification for a GEOSHIELD Application to the SMART Vessel Fluence by a Monte Carlo Simulation

Journal of nuclear science and technology, 2008-06, Vol.45 (sup5), p.24-27

2008

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Ex-Core Detector Response Evaluation of the SMART Reactor by Using the DORT Code

Journal of nuclear science and technology, 2008-06, Vol.45 (sup5), p.78-81

2008

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Radiation Shielding Analysis for the Reactor Assembly of the SMART Reactor

Journal of nuclear science and technology, 2008-06, Vol.45 (sup5), p.82-85

2008

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Spatially dependent embedded self-shielding method for nonuniform temperature distribution

Annals of nuclear energy, 2019-10, Vol.132 (C), p.563-575

2019

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Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

DEVELOPMENT OF MPACT FOR FULL-CORE SIMULATIONS OF MAGNOX GAS-COOLED NUCLEAR REACTORS

A Full-Core Resonance Self-Shielding Method Using a Continuous-Energy Quasi-One-Dimensional Slowing-Down Solution That Accounts for Temperature-Dependent Fuel Subregions and Resonance Interference

Whole Core Transport Calculation Employing Hexagonal Modular Ray Tracing and CMFD Formulation

Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor

Iterative resonance self-shielding methods using resonance integral table in heterogeneous transport lattice calculations

Validation of the SCALE/Polaris−PARCS Code Procedure with the ENDF/B-VII.1 AMPX 56-Group Library: Pressurized Water Reactor

DEVELOPMENT OF THE MPACT 69-GROUP LIBRARY FOR MAGNOX REACTOR ANALYSIS USING CASL VERA

MPACT VERIFICATION WITH MAGNOX REACTOR NEUTRONICS PROGRESSION PROBLEMS

IMPROVEMENT OF SCALE-XSProc MULTIGROUP CROSS SECTION PROCESSING BASED ON THE CENTRM POINTWISE SLOWING DOWN CALCULATION

AN IMPROVED ENERGY DEPOSITION MODEL IN MPACT AND EXPLICIT HEAT GENERATION COUPLING WITH CTF

Gamma transport and diffusion calculation capability coupled with neutron transport simulation in KARMA 1.2

A new procedure to generate resonance integral table with an explicit resonance interference for transport lattice codes

Unstructured partial- and net-current based coarse mesh finite difference acceleration applied to the extended step characteristics method in NEWT

The method of characteristics applied to solving slowing down equation to estimate the self-shielded resonance cross sections with an explicit geometrical effect

Fourier Convergence Analysis of the Rebalance Methods for Discrete Ordinates Transport Equations in Eigenvalue Problems

Axial SPN and Radial MOC Coupled Whole Core Transport Calculation

Development of a computer code system for the analysis of prism and pebble type VHTR cores

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

Development of a physics analysis procedure for the prismatic very high temperature gas-cooled reactors

IAEA GT-MHR benchmark calculations by using the HELIOS/MASTER physics analysis procedure and the MCNP Monte Carlo code

Verification for a GEOSHIELD Application to the SMART Vessel Fluence by a Monte Carlo Simulation

Ex-Core Detector Response Evaluation of the SMART Reactor by Using the DORT Code

Radiation Shielding Analysis for the Reactor Assembly of the SMART Reactor

Spatially dependent embedded self-shielding method for nonuniform temperature distribution

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