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COUPLING OF nTRACER TO COBRA-TF FOR HIGH-FIDELITY ANALYSIS OF VVERs

EPJ Web of conferences, 2021-01, Vol.247, p.2008

2021

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Deuterium permeation and isotope effects in nickel in an elevated temperature range of 450–850 °C

International journal of hydrogen energy, 2014-08, Vol.39 (24), p.12789-12794

2014

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A reduced-basis element method for pin-by-pin reactor core calculations in diffusion and SP3 approximations

Annals of nuclear energy, 2018-06, Vol.116, p.195-209

2018

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Validation of the novel core solver nTRACER/COBRA-TF for full core high fidelity cycle analysis of VVERs

Nuclear engineering and design, 2022-11, Vol.398, p.111946, Article 111946

2022

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Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers

Annals of nuclear energy, 2013-12, Vol.62, p.357-374

2013

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Convergence analysis of fixed-point iteration with Anderson Acceleration on a simplified neutronics/thermal-hydraulics system

Nuclear Engineering and Technology, 2022, 54(2), , pp.532-545

2022

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Analyses of on-the-fly generation of spectral superhomogenization factors for multigroup whole core calculation employing pin-wise slowing-down solutions

Nuclear Engineering and Technology, 2023, 55(3), , pp.1084-1096

2023

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Nodal method for handling irregularly deformed geometries in hexagonal lattice cores

Nuclear Engineering and Technology, 2024, 56(3), , pp.772-784

2024

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Solution of the BEAVRS benchmark using the nTRACER direct whole core calculation code

Journal of nuclear science and technology, 2015-08, Vol.52 (7-8), p.961-969

2015

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Practical methods for GPU-based whole-core Monte Carlo depletion calculation

Nuclear Engineering and Technology, 2023, 55(7), , pp.2516-2533

2023

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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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Development of a drift-flux model based core thermal-hydraulics code for efficient high-fidelity multiphysics calculation

Nuclear engineering and technology, 2019-09, Vol.51 (6), p.1487-1503

2019

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Analysis of C5G7-TD benchmark with a multi-group pin homogenized SP3 code SPHINCS

Nuclear Engineering and Technology, 2021, 53(5), , pp.1403-1415

2021

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Practical resolution of angle dependency of multigroup resonance cross sections using parametrized spectral superhomogenization factors

Nuclear Engineering and Technology, 2017, 49(6), , pp.1287-1300

2017

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High performance 3D pin-by-pin neutron diffusion calculation based on 2D/1D decoupling method for accurate pin power estimation

Nuclear Engineering and Technology, 2021, 53(11), , pp.3543-3562

2021

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RECENT CAPABILITY AND PERFORMANCE ENHANCEMENTS OF THE WHOLE-CORE TRANSPORT CODE nTRACER

EPJ Web of conferences, 2021-01, Vol.247, p.6033

2021

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Special issue on International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering 2017 (M&C 2017)

Nuclear engineering and technology, 2017-09, Vol.49 (6), p.1113-1113

2017

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Development of a fast reactor multigroup cross section generation code EXUS-F capable of direct processing of evaluated nuclear data files

Nuclear Engineering and Technology, 2018, 50(3), , pp.340-355

2018

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Two-Level Coarse Mesh Finite Difference Formulation with Multigroup Source Expansion Nodal Kernels

Journal of nuclear science and technology, 2008-07, Vol.45 (7), p.668-682

2008

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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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PASCAR: Long burning small modular reactor based on natural circulation

Nuclear engineering and design, 2011-05, Vol.241 (5), p.1486-1499

2011

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COARSE MESH FINITE DIFFERENCE ACCELERATION OF DISCRETE ORDINATE NEUTRON TRANSPORT CALCULATION EMPLOYING DISCONTINUOUS FINITE ELEMENT METHOD

Nuclear Engineering and Technology, 2014, 46(6), , pp.783-796

2014

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Analysis of neutron spectrum effects on primary damage in tritium breeding blankets

Journal of nuclear materials, 2012-07, Vol.426 (1-3), p.16-25

2012

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Pointwise cross-section-based on-the-fly resonance interference treatment with intermediate resonance approximation

Nuclear Engineering and Technology, 2015, 47(7), , pp.791-803

2015

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Geometry Extension and Assemblywise Domain Decomposition of nTRACER for Direct Whole-Core Calculation of VVERs

Nuclear science and engineering, 2023-08, Vol.197 (8), p.1564-1583

2023

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COUPLING OF nTRACER TO COBRA-TF FOR HIGH-FIDELITY ANALYSIS OF VVERs

Deuterium permeation and isotope effects in nickel in an elevated temperature range of 450–850 °C

A reduced-basis element method for pin-by-pin reactor core calculations in diffusion and SP3 approximations

Validation of the novel core solver nTRACER/COBRA-TF for full core high fidelity cycle analysis of VVERs

Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers

Convergence analysis of fixed-point iteration with Anderson Acceleration on a simplified neutronics/thermal-hydraulics system

Analyses of on-the-fly generation of spectral superhomogenization factors for multigroup whole core calculation employing pin-wise slowing-down solutions

Nodal method for handling irregularly deformed geometries in hexagonal lattice cores

Solution of the BEAVRS benchmark using the nTRACER direct whole core calculation code

Practical methods for GPU-based whole-core Monte Carlo depletion calculation

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

Development of a drift-flux model based core thermal-hydraulics code for efficient high-fidelity multiphysics calculation

Analysis of C5G7-TD benchmark with a multi-group pin homogenized SP3 code SPHINCS

Practical resolution of angle dependency of multigroup resonance cross sections using parametrized spectral superhomogenization factors

High performance 3D pin-by-pin neutron diffusion calculation based on 2D/1D decoupling method for accurate pin power estimation

RECENT CAPABILITY AND PERFORMANCE ENHANCEMENTS OF THE WHOLE-CORE TRANSPORT CODE nTRACER

Special issue on International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering 2017 (M&C 2017)

Development of a fast reactor multigroup cross section generation code EXUS-F capable of direct processing of evaluated nuclear data files

Two-Level Coarse Mesh Finite Difference Formulation with Multigroup Source Expansion Nodal Kernels

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

PASCAR: Long burning small modular reactor based on natural circulation

COARSE MESH FINITE DIFFERENCE ACCELERATION OF DISCRETE ORDINATE NEUTRON TRANSPORT CALCULATION EMPLOYING DISCONTINUOUS FINITE ELEMENT METHOD

Analysis of neutron spectrum effects on primary damage in tritium breeding blankets

Pointwise cross-section-based on-the-fly resonance interference treatment with intermediate resonance approximation

Geometry Extension and Assemblywise Domain Decomposition of nTRACER for Direct Whole-Core Calculation of VVERs

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