Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL
In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern...
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| Format: | Article |
| Language: | English |
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Wiley
2015-01-01
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| Series: | Science and Technology of Nuclear Installations |
| Online Access: | http://dx.doi.org/10.1155/2015/237646 |
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| author | A. Rais D. Siefman G. Girardin M. Hursin A. Pautz |
| author_facet | A. Rais D. Siefman G. Girardin M. Hursin A. Pautz |
| author_sort | A. Rais |
| collection | DOAJ |
| description | In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor. |
| format | Article |
| id | doaj-art-01c9aef8f61148d7933af2b503180d3f |
| institution | DOAJ |
| issn | 1687-6075 1687-6083 |
| language | English |
| publishDate | 2015-01-01 |
| publisher | Wiley |
| record_format | Article |
| series | Science and Technology of Nuclear Installations |
| spelling | doaj-art-01c9aef8f61148d7933af2b503180d3f2025-08-20T03:22:30ZengWileyScience and Technology of Nuclear Installations1687-60751687-60832015-01-01201510.1155/2015/237646237646Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPLA. Rais0D. Siefman1G. Girardin2M. Hursin3A. Pautz4École Polytechnique Fédérale de Lausanne (EPFL), 1015 Lausanne, SwitzerlandÉcole Polytechnique Fédérale de Lausanne (EPFL), 1015 Lausanne, SwitzerlandÉcole Polytechnique Fédérale de Lausanne (EPFL), 1015 Lausanne, SwitzerlandPaul Scherrer Institut (PSI), 5232 Villigen, SwitzerlandÉcole Polytechnique Fédérale de Lausanne (EPFL), 1015 Lausanne, SwitzerlandIn order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.http://dx.doi.org/10.1155/2015/237646 |
| spellingShingle | A. Rais D. Siefman G. Girardin M. Hursin A. Pautz Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL Science and Technology of Nuclear Installations |
| title | Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL |
| title_full | Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL |
| title_fullStr | Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL |
| title_full_unstemmed | Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL |
| title_short | Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL |
| title_sort | methods and models for the coupled neutronics and thermal hydraulics analysis of the crocus reactor at efpl |
| url | http://dx.doi.org/10.1155/2015/237646 |
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