Results 151 to 160 of about 2,762 (182)
Some of the next articles are maybe not open access.
Annals of Nuclear Energy, 2020
Abstract SN nodal method has been successfully applied to fast reactor core neutronics analysis. The computational efficiency can be enhanced efficiently by the popular coarse-mesh finite difference (CMFD) method. However, the CMFD method suffers from the problem of instability, especially for optically thick problems.
Youqi Zheng, Hongchun Wu
exaly +2 more sources
Abstract SN nodal method has been successfully applied to fast reactor core neutronics analysis. The computational efficiency can be enhanced efficiently by the popular coarse-mesh finite difference (CMFD) method. However, the CMFD method suffers from the problem of instability, especially for optically thick problems.
Youqi Zheng, Hongchun Wu
exaly +2 more sources
NEMSQR: A 3-D multi group diffusion theory code based on nodal expansion method for square geometry
Annals of Nuclear Energy, 2014Abstract A three dimensional, multigroup, neutron diffusion theory based computer code NEMSQR ( N odal E xpansion M ethod for SQ uare geometry) is developed for square geometry in order to perform reactor core calculation. The code is based on Nodal Expansion Method (NEM).
Tanay Mazumdar +2 more
exaly +2 more sources
Kerntechnik, 2015
Abstract In this paper, we developed a new approach of analytic function expansion nodal (AFEN) method to solve the multi-group and multi-dimensional neutron diffusion equation in reactor cores with hexagonal fuel assembly. This method represents a multidimensional intra nodal flux distribution in terms of analytic basis functions at any
A Pazirandeh
exaly +2 more sources
Abstract In this paper, we developed a new approach of analytic function expansion nodal (AFEN) method to solve the multi-group and multi-dimensional neutron diffusion equation in reactor cores with hexagonal fuel assembly. This method represents a multidimensional intra nodal flux distribution in terms of analytic basis functions at any
A Pazirandeh
exaly +2 more sources
Kerntechnik, 2019
Abstract The C-PORCA/HELIOS models have been used at NPP Paks as basic core neutron physics calculation tools for many years. C-PORCA is a node-wise diffusion model for the purpose of 3D core analysis. HELIOS is a well-known neutron transport code. Its utilisation at Paks NPP has a dual use.
I. Pós +4 more
exaly +2 more sources
Abstract The C-PORCA/HELIOS models have been used at NPP Paks as basic core neutron physics calculation tools for many years. C-PORCA is a node-wise diffusion model for the purpose of 3D core analysis. HELIOS is a well-known neutron transport code. Its utilisation at Paks NPP has a dual use.
I. Pós +4 more
exaly +2 more sources
Annals of Nuclear Energy, 2021
Abstract The rod ejection accident (REA) is the most severe reactivity-initiated accident of the pressurized water reactors (PWRs). The three-dimensional (3-D) neutron kinetics method is widely applied to the REA analysis, where the fixed neutron source is usually ignored at the hot zero power (HZP) condition. This paper describes a new capability of
YuGwon Jo, Ho Cheol Shin
openaire +1 more source
Abstract The rod ejection accident (REA) is the most severe reactivity-initiated accident of the pressurized water reactors (PWRs). The three-dimensional (3-D) neutron kinetics method is widely applied to the REA analysis, where the fixed neutron source is usually ignored at the hot zero power (HZP) condition. This paper describes a new capability of
YuGwon Jo, Ho Cheol Shin
openaire +1 more source
Kerntechnik, 2017
Abstract Continuous-energy Monte Carlo reactor physics code Serpent 2 was used to model the critical steady state conditions measured in V-1000 zero-power critical facility at Kurchatov Institute (KI), Moscow in 1990–1992. The Serpent 2 results were compared to measurements and Serpent 2 was used to generate group constants for reactor ...
Sahlberg, Ville
openaire +4 more sources
Abstract Continuous-energy Monte Carlo reactor physics code Serpent 2 was used to model the critical steady state conditions measured in V-1000 zero-power critical facility at Kurchatov Institute (KI), Moscow in 1990–1992. The Serpent 2 results were compared to measurements and Serpent 2 was used to generate group constants for reactor ...
Sahlberg, Ville
openaire +4 more sources
Analytic Error Analysis of Cross Section Interpolation Methods in Nodal Diffusion Codes-I: Theory
International Conference on Physics of Reactors 2022 (PHYSOR 2022), 2022Thomas Folk +4 more
openaire +1 more source
Analytic Reactivity Error Analysis of Cross Section Interpolation Methods in Nodal Diffusion Codes
Transactions of the American Nuclear Society, 2022Thomas Folk +2 more
openaire +1 more source
Development of Rast-Ai: A Hybrid Nodal Diffusion/Artificial Neural Network Code for Reactor Analysis
SSRN Electronic Journal, 2022Siarhei Dzianisau +2 more
openaire +1 more source