Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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The applicability study and validation of TULIP code for full energy range spectrum

  • Wenjie Chen (School of Nuclear Science and Technology, Xi'an Jiaotong University) ;
  • Xianan Du (School of Nuclear Science and Technology, Xi'an Jiaotong University) ;
  • Rong Wang (School of Nuclear Science and Technology, Xi'an Jiaotong University) ;
  • Youqi Zheng (School of Nuclear Science and Technology, Xi'an Jiaotong University) ;
  • Yongping Wang (School of Nuclear Science and Technology, Xi'an Jiaotong University) ;
  • Hongchun Wu (School of Nuclear Science and Technology, Xi'an Jiaotong University)
  • Received : 2023.07.19
  • Accepted : 2023.08.24
  • Published : 2023.12.25
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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 neutron-moderating 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.

Keywords

Acknowledgement

This work is supported by National Natural Science Foundation of China (No.12005160).

References

  1. W.S. Yang, Fast reactor physics and computational methods, Nucl. Eng. Technol. 44 (2) (2012) 177-198, https://doi.org/10.5516/net.01.2012.504. 
  2. C. Lee, W.S. Yang, MC2-3: multigroup cross section generation code for fast reactor analysis, Nucl. Sci. Eng. 187 (3) (2017) 268-290, https://doi.org/10.1080/00295639.2017.1320893. 
  3. J.M. Ruggieri, J. Tommasi, J.F. Lebrat, et al., Eranos 2.1: international code system for GEN IV fast reactor analysis, Proceedings of ICAPP (2006) 2432-2439. 
  4. G. Rimpault, D. Plisson, J. Tommasi, et al., The ERANOS code and data system for fast reactor neutronic analyses, in: PHYSOR 2002-International Conference on the New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing, 2002. 
  5. H. Guo, E. Garcia, B. Faure, et al., Advanced method for neutronic simulation of control rods in sodium fast reactors: numerical and experimental validation, Ann. Nucl. Energy 129 (2019) 90-100, https://doi.org/10.1016/j.anucene.2019.01.042.
  6. B. Roque, A. Rizzo, V. Pascal, et al., Experimental validation of the new code package APOLLO3-SFR against ZPPR-10A experiment for critical and voided configurations//PHYSOR 2016, in: International Conference on the Advances in Reactor Physics Unifying Theory and Experiments in the 21st Century, 2016. 
  7. T. Hazama, G. Chiba, K. Sugino, Development of a fine and ultra-fine group cell calculation code SLAROM-UF for fast reactor analyses, J. Nucl. Sci. Technol. 43 (8) (2006) 908-918, https://doi.org/10.1080/18811248.2006.9711176 ([LinkOut]). 
  8. Bae M H, Choi Y W, Shin A, et al. Preliminary Development of SFR Nuclear Code System for Regulatory Evaluation.//Proceedings of the KNS 2016 Autumn Meeting. 
  9. C. Lim, H.G. Joo, W.S. Yang, Development of a fast reactor multigroup cross section generation code EXUS-F capable of direct processing of evaluated nuclear data files, Nucl. Eng. Technol. 50 (3) (2018) 340-355, https://doi.org/10.1016/j.net.2018.01.013. 
  10. Xiaoyan Yang, Hong Yu, Zhi Gang, et al., Loading scheme research on the first criticality of China experimental fast reactor, Atomic Energy Sci. Technol. 47 (S1) (2013) 58-61, https://doi.org/10.7538/yzk.2013.47.S0.0058 (In Chinese). 
  11. K. Hu, X.B. Ma, T. Zhang, et al., MGGC2.0: a preprocessing code for the multi-group cross section of the fast reactor with ultrafine group library, Nucl. Eng. Technol. 55 (8) (2023) 2785-2796, https://doi.org/10.1016/j.net.2023.05.005. 
  12. T. Zhang, Z. Li, Variational nodal methods for neutron transport: 40 years in review, Nucl. Eng. Technol. (2022 Apr 27) 3181-3204.  https://doi.org/10.1016/j.net.2022.04.012
  13. T. Zhang, W. Xiao, H. Yin, Q. Sun, X. Liu, VITAS: a multi-purpose simulation code for the solution of neutron transport problems based on variational nodal methods, Ann. Nucl. Energy (2022 July 178), 109335. 
  14. X. Du, L. Cao, Y. Zheng, et al., A hybrid method to generate few-group cross sections for fast reactor analysis, J. Nucl. Sci. Technol. 55 (8) (2018) 931-944, https://doi.org/10.1080/00223131.2018.1452650. 
  15. S.C. Zhou, H.C. Wu, L. Cao, et al., LAVENDER: a steady-state core analysis code for design studies of accelerator driven subcritical reactors, Nucl. Eng. Des. 278 (2014) 434-444, https://doi.org/10.1016/j.nucengdes.2014.07.027. 
  16. M. He, H. Wu, Y. Zheng, et al., Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method, Nucl. Eng. Des. 295 (2015) 489-499, https://doi.org/10.1016/j.nucengdes.2015.10.021. 
  17. Y. Zheng, L. Qiao, Z. Zhai, et al., SARAX: a new code for fast reactor analysis part II: verification, validation and uncertainty quantification, Nucl. Eng. Des. 331 (2018) 41-53, https://doi.org/10.1016/j.nucengdes.2018.02.033. 
  18. X. Du, J. Choe, T. Tran, et al., Neutronic simulation of China Experimental Fast Reactor start-up tests. Part I: SARAX code deterministic calculation, Ann. Nucl. Energy 136 (2020), 107046, https://doi.org/10.1016/j.anucene.2019.107046. 
  19. H.D.L. Mencarini, C.J. King, Fuel geometry options for a moderated low-enriched uranium kilowatt-class space nuclear reactor, Nucl. Eng. Des. 340 (2018), https://doi.org/10.1016/j.nucengdes.2018.09.017. 
  20. B.K. Jeon, W.S. Yang, Y.S. Jung, et al., in: Extension of MC2-3 for Generation of Multigroup Cross Sections in Thermal Energy range//Proceedings of M&C 2017 Conference, Jeju, Korea, 2017, https://doi.org/10.1016/B978-0-08-019610-7.50043-9. 
  21. K. Kim, M.L. Williams, A.M. Holcomb, et al., The SCALE/AMPX multigroup cross section processing for fast reactor analysis, Ann. Nucl. Energy 132 (2019) 161-171, https://doi.org/10.1016/j.anucene.2019.04.034. 
  22. L. Wei, Y. Zheng, X. Du, et al., Extension of SARAX code system for reactors with intermediate spectrum, Nucl. Eng. Des. (2020) 370, https://doi.org/10.1016/j.nucengdes.2020.110883. 
  23. X. DU, Few-Group Cross Section Generation Method for Fast Reactor Analysis and its Verification and Validation. Doctoral Dissertation, Xi 'an Jiaotong University, Xi 'an, 2018 (In Chinese). 
  24. J.B. Briggs, L. Scott, A. Nouri, The international criticality safety benchmark evaluation project, Nucl. Sci. Eng. 145 (1) (2003) 1-10, https://doi.org/10.13182/nse03-14. 
  25. C. Lee, Y. Jung, W. Yang, MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis Nuclear, Argonne National Lab.(ANL), Argonne, IL (United States), 2018. 
  26. C. Perey, F. Perey, J. Harvey, et al., 56Fe Resonance Parameters for Neutron Energies up to 850 keV, Oak Ridge National Lab, 1990, https://doi.org/10.2172/6028649. 
  27. M. Chadwick, P. Oblozinsky, M. Herman, et al., ENDF/B-VII.0: next generation evaluated nuclear data library for nuclear science and technology, Nucl. Data Sheets 107 (12) (2006), https://doi.org/10.1016/j.nds.2006.11.001. 
  28. T. Zu, J. Xu, Y. Tang, et al., NECP-Atlas: a new nuclear data processing code, Ann. Nucl. Energy 123 (2019), https://doi.org/10.1016/j.anucene.2018.09.016. 
  29. Q. He, Q. Zheng, J. Li, et al., NECP-MCX: a hybrid Monte-Carlo-Deterministic particle-transport code for the simulation of deep-penetration problems, Ann. Nucl. Energy (2021) 151, https://doi.org/10.1016/j.anucene.2020.107978. 
  30. Q. Zhang, H. Wu, L. Cao, Y. Zheng, An improved resonance self-shielding calculation method based on equivalence theory, Nucl. Sci. Eng. 179 (2015) 233-252, https://doi.org/10.13182/NSE13-108.