Neutronic simulation of the CEFR experiments with the nodal diffusion code system RAST-F
Nuclear Engineering and Technology, 2022 CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect.
Tuan Quoc Tran, Deokjung Lee
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Development of nodal diffusion code RAST-V for Vodo–Vodyanoi Energetichesky reactor analysis
Nuclear Engineering and Technology, 2022 This paper presents the development of a nodal diffusion code, RAST-V, and its verification and validation for VVER (vodo–vodyanoi energetichesky reactor) analysis.
Jaerim Jang +2 more
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RAST-K v2—Three-Dimensional Nodal Diffusion Code for Pressurized Water Reactor Core Analysis
Energies, 2020 The RAST-K v2, a novel nodal diffusion code, was developed at the Ulsan National Institute of Science and Technology (UNIST) for designing the cores of pressurized water reactors (PWR) and performing analyses with high accuracy and computational ...
Jinsu Park +7 more
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Nuclear Engineering and TechnologyThis study presents the uncertainty quantification results of steady-state depletion simulations using a multi-physics coupling framework based on the nodal diffusion code RAST-K. Developed for the analysis and optimization of pressurized water reactors,
Jinsu Park, Yeongseok Kang, Deokjung Lee
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New 3D Diffusion Code Based on The Nodal Polynomial Expansion
Brazilian Journal of Radiation Sciences, 2021 Nodal Expansion Method (NEM) is widely employed in the neutronic design of nuclear reactors core. The main reason is its higher computational performance and efficiency when compared with the conventional fine mesh difference method or finite element ...
Jadna Mara Santos Mendes +2 more
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Three-Dimensional Multigroup Diffusion Code ANDEX Based on Nodal Method for Cartesian Geometry [PDF]
Journal of Nuclear Science and Technology, 1990 An analytic polynomial nodal method using partial currents has been derived for the solution of multigroup neutron diffusion equations in three-dimensional (3-D) cartesian geometry. This method is characterized by expressing the source and leakage terms in an auxiliary 1-D diffusion equation by quadratic polynomials and solving it analytically.
Toshikazu Takeda
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ANALYSIS OF THE OECD/NEA SFR BENCHMARK WITH ANTS REDUCED-ORDER NODAL DIFFUSION SOLVER AND THE SERPENT MONTE CARLO CODE [PDF]
EPJ Web of Conferences, 2021 In order to meet modern industrial and scientific demands the Kraken multi-physics platform’s development was recently launched at VTT Technical Research Centre of Finland.
Szogradi Marton
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Development and Verification of Hexagonal Quasi-diffusion Code Based on Variational Nodal Method
Yuanzineng kexue jishu, 2023 New type reactors usually adopt advanced fuel types and more anisotropic fuel arrangements, the neutron spectrum is very different from that of pressurized water reactor, and the neutron anisotropy is stronger.
ZHUANG Kun;WANG Lianjie;LIU Kun;YAN Jiangtao;SHANG Wen
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VERIFICATION AND VALIDATION OF BACK-END CYCLE SOURCE TERM CALCULATION OF THE NODAL CODE RAST-K [PDF]
EPJ Web of Conferences, 2021 Verification and validation (V&V) results of source term calculation capability implemented in the nodal diffusion code RAST-K are presented in this paper.
Jang Jaerim +4 more
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VERIFICATION OF THE DIFFUSION AND TRANSPORT SOLVERS WITHIN DIF3D FOR 3D HEXAGONAL GEOMETRIES [PDF]
EPJ Web of Conferences, 2021 The DIF3D code (DIFfusion 3D) has been a workhorse of fast reactor analysis work at Argonne National Laboratory for over 40 years. DIF3D was primarily built in the late 1970s as a three-dimensional multigroup diffusion equation solver operating on semi ...
Nelson A. G., Smith M. A., Heidet F.
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