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Volume 45 Issue S1
Jun.  2024
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Sun Lin, Wu Zongyun, Zhang Zhenyu, Xue Fangyuan, Wang Xuesong, Liu Tiancai. Development of System Analysis Code for Sodium-cooled Fast Reactor and its Verification on SHRT-45R Benchmark Problem[J]. Nuclear Power Engineering, 2024, 45(S1): 63-67. doi: 10.13832/j.jnpe.2024.S1.0063
Citation: Sun Lin, Wu Zongyun, Zhang Zhenyu, Xue Fangyuan, Wang Xuesong, Liu Tiancai. Development of System Analysis Code for Sodium-cooled Fast Reactor and its Verification on SHRT-45R Benchmark Problem[J]. Nuclear Power Engineering, 2024, 45(S1): 63-67. doi: 10.13832/j.jnpe.2024.S1.0063

Development of System Analysis Code for Sodium-cooled Fast Reactor and its Verification on SHRT-45R Benchmark Problem

doi: 10.13832/j.jnpe.2024.S1.0063
  • Received Date: 2023-10-21
  • Rev Recd Date: 2024-01-01
  • Publish Date: 2024-06-15
  • The system analysis software for sodium-cooled fast reactors based on best estimation methods is crucial for both reactor system design and safety review. In this study, the Fast Reactor Transient Analysis Code(FRTAC) of sodium-cooled fast reactor system suitable for general reactor type is developed. Besides traditional input card, a graphical user interface is added. Through internal testing and third-party testing, FRTAC simulation function can cover the normal operation, transient operation, design basis conditions and some design extended conditions. In order to verify the accuracy of the software, the International Atomic Energy Agency SHRT-45R benchmark problems were used for modeling and analysis. The results prove that the errors of key parameters such as core coolant temperature and flow rate between calculation results and experimental values were less than 10%. The FRTAC calculation results of the software are accurate and can be used for accident analysis of sodium-cooled fast reactor.

     

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  • [1]
    DEL NEVO A, MARTELLI E. Validation of a three-dimensional model of EBR-II and assessment of RELAP5-3D based on SHRT-17 test[J]. Nuclear Technology, 2016, 193(1): 1-14. doi: 10.13182/NT14-152
    [2]
    ROELOFS F, UITSLAG-DOOLAARD H, ZWIJSEN K, et al. Fast reactor thermal hydraulics in the Dutch PIONEER program[J]. Nuclear Engineering and Design, 2023, 412: 112473. doi: 10.1016/j.nucengdes.2023.112473
    [3]
    FANNING T, BRUNETT A, SUMNER T. The SAS4A/SASSYS-1 safety analysis code system, Version 5: ANL/NE-16/19[R]. Argonne: Argonne National Lab, 2017.
    [4]
    HU R, ZOU L, HU G J. SAM user’s guide: ANL/NSE-19/18[R]. Argonne: Argonne National Lab, 2019.
    [5]
    王晋,张东辉. 快堆系统分析程序FASYS堆芯分析模块验证[J]. 原子能科学技术,2020, 54(2): 264-272. doi: 10.7538/yzk.2019.youxian.0086
    [6]
    王晓坤,齐少璞,杨军,等. 钠冷快堆系统程序FR-Sdaso开发[J]. 原子能科学技术,2020, 54(11): 2045-2053. doi: 10.7538/yzk.2020.youxian.0306
    [7]
    隋丹婷,陆道纲,郭超. EBR-Ⅱ余热排出实验及非能动余热排出系统性能分析[J]. 原子能科学技术,2018, 52(5): 881-890. doi: 10.7538/yzk.2018.52.05.0881
    [8]
    MA Z Y, YUE N N, ZHENG M Y, et al. Basic verification of THACS for sodium-cooled fast reactor system analysis[J]. Annals of Nuclear Energy, 2015, 76: 1-11. doi: 10.1016/j.anucene.2014.09.025
    [9]
    SUMNER T S, WEI T Y C. Benchmark specifications and data requirements for EBR-II shutdown heat removal tests SHRT-17 and SHRT-45R: ANL-ARC-226 Rev. 1[R]. Argonne: Argonne National Lab, 2012.
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