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
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 c...
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| Format: | Article |
| Language: | English |
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Brazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR)
2022-02-01
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| Series: | Brazilian Journal of Radiation Sciences |
| Subjects: | |
| Online Access: | https://bjrs.org.br/revista/index.php/REVISTA/article/view/1738 |
| _version_ | 1797985911894441984 |
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| author | Pedro Henrique Silva Rodrigues José Rubens Maiorino Roberto Asano Júnior |
| author_facet | Pedro Henrique Silva Rodrigues José Rubens Maiorino Roberto Asano Júnior |
| author_sort | Pedro Henrique Silva Rodrigues |
| collection | DOAJ |
| description |
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 SCALE 6.0, 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 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.
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| first_indexed | 2024-04-11T07:25:20Z |
| format | Article |
| id | doaj.art-53a9a4face8b4a4a853e381f0d8b914b |
| institution | Directory Open Access Journal |
| issn | 2319-0612 |
| language | English |
| last_indexed | 2024-04-11T07:25:20Z |
| publishDate | 2022-02-01 |
| publisher | Brazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR) |
| record_format | Article |
| series | Brazilian Journal of Radiation Sciences |
| spelling | doaj.art-53a9a4face8b4a4a853e381f0d8b914b2022-12-22T04:37:05ZengBrazilian Radiation Protection Society (Sociedade Brasileira de Proteção Radiológica, SBPR)Brazilian Journal of Radiation Sciences2319-06122022-02-0110110.15392/bjrs.v10i1.1738On development of in-core fuel system for PWR reactors: Part I generation of macroscopic cross sections using SCALE 6.0 for use in nodal calculationPedro Henrique Silva Rodrigues0José Rubens Maiorino1Roberto Asano Júnior2Universidade Federal do ABCUniversidade Federal do ABCUniversidade Federal do ABC 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 SCALE 6.0, 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 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. https://bjrs.org.br/revista/index.php/REVISTA/article/view/1738in-core fuel managementPWRbenchmarkSCALE |
| spellingShingle | Pedro Henrique Silva Rodrigues José Rubens Maiorino Roberto Asano Júnior 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 Brazilian Journal of Radiation Sciences in-core fuel management PWR benchmark SCALE |
| 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 |
| title_full | 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 |
| title_fullStr | 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 |
| title_full_unstemmed | 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 |
| title_short | 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 |
| title_sort | 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 | in-core fuel management PWR benchmark SCALE |
| url | https://bjrs.org.br/revista/index.php/REVISTA/article/view/1738 |
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