The applicability study and validation of TULIP code for full energy range spectrum
0
조회
0
다운로드
0
인용
0
참고문헌
키워드
초록
NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi’an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutronmoderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system
Abstract
NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi’an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutronmoderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system
참고문헌 (30)
- W. S. YANG, "FAST REACTOR PHYSICS AND COMPUTATIONAL METHODS", Nuclear Engineering and Technology, 44(2), pp.177~198 (2012), 한국원자력학회
- Changho Lee, "MC<sup>2<", sup>-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis, 2017, pp.187(3), 268~290[DOI]
- J.M. Ruggieri, "Eranos 2.1: international code system for GEN IV fast reactor analysis", Proceedings of ICAPP, pp.2432~2439 (2006)
- G. Rimpault, "The ERANOS code and data system for fast reactor neutronic analyses", PHYSOR 2002-International Conference on the New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing (2002)
- H. Guo, "Advanced method for neutronic simulation of control rods in sodium fast reactors: Numerical and experimental validation", Annals of Nuclear Energy, 129, pp.90~100 (2019), Elsevier BV[DOI]
- B. Roque, "Experimental validation of the new code package APOLLO3-SFR against ZPPR-10A experiment for critical and voided configurations", International Conference on the Advances in Reactor Physics Unifying Theory and Experiments in the 21st Century, pp.2016[DOI]
- Taira HAZAMA, "Development of a Fine and Ultra-Fine Group Cell Calculation Code SLAROM-UF for Fast Reactor Analyses", Journal of Nuclear Science and Technology, 43(8), pp.908~918 (2006), Informa UK Limited[DOI]
- Bae M H, "Preliminary Development of SFR Nuclear Code System for Regulatory Evaluation", Proceedings of the KNS 2016 Autumn Meeting (2016)
- 임창현, "Development of a fast reactor multigroup cross section generation code EXUS-F capable of direct processing of evaluated nuclear data files", Nuclear Engineering and Technology, 50(3), pp.340~355 (2018), 한국원자력학회
- Xiaoyan Yang, "Loading scheme research on the first criticality of China experimental fast reactor", Atomic Energy Sci. Technol, 47(S1), pp.58~61 (2013)[DOI]
- Hu Kui, "MGGC2.0: A preprocessing code for the multi-group cross section of the fast reactor with ultrafine group library", Nuclear Engineering and Technology, 55(8), pp.2785~2796 (2023), 한국원자력학회[DOI]
- Zhang Tengfei, "Variational nodal methods for neutron transport: 40 years in review", Nuclear Engineering and Technology, 54(9), pp.3181~3204 (2022), 한국원자력학회[DOI]
- T. Zhang, "VITAS: a multi-purpose simulation code for the solution of neutron transport problems based on variational nodal methods", Ann. Nucl. Energy, 178, pp.109335~ (2022)
- Xianan Du, "A hybrid method to generate few-group cross sections for fast reactor analysis", Journal of Nuclear Science and Technology, 55(8), pp.931~944 (2018), Informa UK Limited[DOI]
- Shengcheng Zhou, "LAVENDER: A steady-state core analysis code for design studies of accelerator driven subcritical reactors", Nuclear Engineering and Design, 278, pp.434~444 (2014), Elsevier BV[DOI]
- Mingtao He, "Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method", Nuclear Engineering and Design, 295, pp.489~499 (2015), Elsevier BV[DOI]
- Youqi Zheng, "SARAX: A new code for fast reactor analysis part II: Verification, validation and uncertainty quantification", Nuclear Engineering and Design, 331, pp.41~53 (2018), Elsevier BV[DOI]
- Xianan Du, "Neutronic simulation of China Experimental Fast Reactor start-up tests. Part I: SARAX code deterministic calculation", Annals of Nuclear Energy, 136, pp.107046~ (2020), Elsevier BV[DOI]
- Leonardo de Holanda Mencarini, "Fuel geometry options for a moderated low-enriched uranium kilowatt-class space nuclear reactor", Nuclear Engineering and Design, 340, pp.122~132 (2018), Elsevier BV[DOI]
- B.K. Jeon, "Extension of MC2-3 for Generation of Multigroup Cross Sections in Thermal Energy range", Proceedings of M&C 2017 Conference (2017)
- Kang Seog Kim, "The SCALE", AMPX multigroup cross section processing for fast reactor analysis, 2019, pp.132, 161~171[DOI]
- Linfang Wei, "Extension of SARAX code system for reactors with intermediate spectrum", Nuclear Engineering and Design, 370, pp.110883~ (2020), Elsevier BV[DOI]
- X. DU, "Few-Group Cross Section Generation Method for Fast Reactor Analysis and its Verification and Validation", Doctor, pp.2018
- J. Blair Briggs, "The International Criticality Safety Benchmark Evaluation Project", Nuclear Science and Engineering, 145(1), pp.1~10 (2003), Informa UK Limited[DOI]
- MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis Nuclear, "C. Lee", Argonne National Lab.(ANL), ~ (2018)
- 56Fe Resonance Parameters for Neutron Energies up to 850 keV, "C. Perey", Oak Ridge National Lab, ~ (1990)
- M.B. Chadwick, "ENDF", B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology, 2006, pp.107(12), 2931~3060[DOI]
- Tiejun Zu, "NECP-Atlas: A new nuclear data processing code", Annals of Nuclear Energy, 123, pp.153~161 (2019), Elsevier BV[DOI]
- Qingming He, "NECP-MCX: A hybrid Monte-Carlo-Deterministic particle-transport code for the simulation of deep-penetration problems", Annals of Nuclear Energy, 151, pp.107978~ (2021), Elsevier BV[DOI]
- Qian Zhang, "An Improved Resonance Self-Shielding Calculation Method Based on Equivalence Theory", Nuclear Science and Engineering, 179(3), pp.233~252 (2015), Informa UK Limited[DOI]