Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems
A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a t...
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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/10.1155/2015/180979 |
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| author | Wonkyeong Kim Jinsu Park Tomasz Kozlowski Hyun Chul Lee Deokjung Lee |
| author_facet | Wonkyeong Kim Jinsu Park Tomasz Kozlowski Hyun Chul Lee Deokjung Lee |
| author_sort | Wonkyeong Kim |
| collection | DOAJ |
| description | A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2. |
| format | Article |
| id | doaj-art-1071b93140a047c0aeca31964bc9ead3 |
| institution | Kabale University |
| 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-1071b93140a047c0aeca31964bc9ead32025-08-20T03:55:45ZengWileyScience and Technology of Nuclear Installations1687-60751687-60832015-01-01201510.1155/2015/180979180979Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code SystemsWonkyeong Kim0Jinsu Park1Tomasz Kozlowski2Hyun Chul Lee3Deokjung Lee4Ulsan National Institute of Science and Technology, UNIST Gil-50, Ulsan 689-798, Republic of KoreaUlsan National Institute of Science and Technology, UNIST Gil-50, Ulsan 689-798, Republic of KoreaUniversity of Illinois, Urbana-Champaign, IL 61801, USAKorea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353, Republic of KoreaUlsan National Institute of Science and Technology, UNIST Gil-50, Ulsan 689-798, Republic of KoreaA high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.http://dx.doi.org/10.1155/2015/180979 |
| spellingShingle | Wonkyeong Kim Jinsu Park Tomasz Kozlowski Hyun Chul Lee Deokjung Lee Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems Science and Technology of Nuclear Installations |
| title | Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems |
| title_full | Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems |
| title_fullStr | Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems |
| title_full_unstemmed | Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems |
| title_short | Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems |
| title_sort | comparative neutronics analysis of dimple s06 criticality benchmark with contemporary reactor core analysis computer code systems |
| url | http://dx.doi.org/10.1155/2015/180979 |
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