On development of in-core fuel system for PWR reactors: Part I generation of macroscopic cross sections using SCALE 6.0 for use in nodal calculation
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| Format: | Artículo científico |
| Langue: | en |
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Sociedade Brasileira de Proteção Radiológica - SBPR
2022
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| _version_ | 1876441105583570944 |
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| author | P.H.S Rodrigues |
| author_facet | P.H.S Rodrigues |
| contents | On development of in-core fuel system for PWR reactors: Part I generation of macroscopic cross sections using SCALE 6.0 for use in nodal calculation P.H.S Rodrigues J.R Maiorino R. JR. Asano Biología In PWR SCALE benchmark core fuel management This work presents the description of the first part of a methodology applied to perform In-Core Fuel Management (ICFM) in Pressurized Water Reactor (PWR). The ICFM of a PWR reactor consists on defining the best charging or recharging pattern of fuel assemblies inside a reactor for an operational cycle. This means, finding a suitable arrangement of fuel assemblies that optimizes the performance of the reactor, which complies with all safety criteria. Genetic algorithms (GAs) are used to select the arrangements that interact with the reactor physics simulation code, holding the neutron characteristics of each fuel assembly. Therefore, a reliable and fast code was developed accordingly. The consolidated technique of coarse mesh node code that numerically solves the multigroup diffusion equation for two groups of energy, fast and thermal neutrons, in two dimensions was selected. In this type of code, it is essential that each fuel assembly is homogenized and characterized by its macroscopic cross sections, for each reactor’s burnup condition. The cross sections are generated with the support of NEWT, ORIGEN and TRITON modules in SCALE 6.0, a computational platform developed by the Reactor and Nuclear Systems Division (RNSD), from the Oak Ridge National Laboratory (ORNL). The completeness of the qualification and validation of the results obtained from the homogenization of the fuel assembly by the SCALE was performed comparing the infinity multiplication factor results with actual data of a benchmark reactor. The fully documented Almaraz Nuclear Power Plant provided by the International Atomic Energy Agency (IAEA)-TECDOC-815, has been used as benchmark with successful results. 2022 artículo científico 2319-0612 https://www.redalyc.org/articulo.oa?id=722280971008 en http://www.redalyc.org/revista.oa?id=7222 Brazilian Journal of Radiation Sciences application/pdf Sociedade Brasileira de Proteção Radiológica - SBPR Brazilian Journal of Radiation Sciences (Brasil) Num.1 Vol.10 |
| format | Artículo científico |
| id | redalyc_722280971008 |
| institution | Redalyc |
| language | en |
| publishDate | 2022 |
| publisher | Sociedade Brasileira de Proteção Radiológica - SBPR |
| spellingShingle | On development of in-core fuel system for PWR reactors: Part I generation of macroscopic cross sections using SCALE 6.0 for use in nodal calculation P.H.S Rodrigues Biología In PWR SCALE benchmark core fuel management On development of in-core fuel system for PWR reactors: Part I generation of macroscopic cross sections using SCALE 6.0 for use in nodal calculation P.H.S Rodrigues J.R Maiorino R. JR. Asano Biología In PWR SCALE benchmark core fuel management This work presents the description of the first part of a methodology applied to perform In-Core Fuel Management (ICFM) in Pressurized Water Reactor (PWR). The ICFM of a PWR reactor consists on defining the best charging or recharging pattern of fuel assemblies inside a reactor for an operational cycle. This means, finding a suitable arrangement of fuel assemblies that optimizes the performance of the reactor, which complies with all safety criteria. Genetic algorithms (GAs) are used to select the arrangements that interact with the reactor physics simulation code, holding the neutron characteristics of each fuel assembly. Therefore, a reliable and fast code was developed accordingly. The consolidated technique of coarse mesh node code that numerically solves the multigroup diffusion equation for two groups of energy, fast and thermal neutrons, in two dimensions was selected. In this type of code, it is essential that each fuel assembly is homogenized and characterized by its macroscopic cross sections, for each reactor’s burnup condition. The cross sections are generated with the support of NEWT, ORIGEN and TRITON modules in SCALE 6.0, a computational platform developed by the Reactor and Nuclear Systems Division (RNSD), from the Oak Ridge National Laboratory (ORNL). The completeness of the qualification and validation of the results obtained from the homogenization of the fuel assembly by the SCALE was performed comparing the infinity multiplication factor results with actual data of a benchmark reactor. The fully documented Almaraz Nuclear Power Plant provided by the International Atomic Energy Agency (IAEA)-TECDOC-815, has been used as benchmark with successful results. 2022 artículo científico 2319-0612 https://www.redalyc.org/articulo.oa?id=722280971008 en http://www.redalyc.org/revista.oa?id=7222 Brazilian Journal of Radiation Sciences application/pdf Sociedade Brasileira de Proteção Radiológica - SBPR Brazilian Journal of Radiation Sciences (Brasil) Num.1 Vol.10 |
| title | On development of in-core fuel system for PWR reactors: Part I generation of macroscopic cross sections using SCALE 6.0 for use in nodal calculation |
| topic | Biología In PWR SCALE benchmark core fuel management |
| url | https://www.redalyc.org/articulo.oa?id=722280971008 |