SARAX程序系统在钠冷快堆瞬态分析中的应用
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摘要
无保护事故下的瞬态分析是钠冷快堆安全分析的重要内容。基于OECD/NEA发布的MOX-3600和MET-1000基准题,本文利用SARAX程序系统对不同钠冷快堆进行了瞬态计算,分析了堆内各种反应性反馈效应,并计算了无保护失流(ULOF)事故和无保护超功率运行(UTOP)事故下燃料温度和冷却剂温度的变化。计算结果表明:SARAX程序系统在快堆瞬态分析中可给出合理的参数预测结果;ULOF事故对于钠冷快堆是更为严重的事故瞬态,会导致堆内的钠沸腾进而发生严重事故。
The transient analysis of unprotected accident is significant in safety analysis of sodium-cooled fast reactor. Based on MOX-3600 and MET-1000 benchmarks published by OECD/NEA, SARAX code system was applied to do transient calculation for different sodium-cooled fast reactors. Various reactivity feedback effects in the reactor were analyzed, and the changes of fuel temperature and coolant temperature during unprotected loss of flow(ULOF) transient and unprotected transient over power(UTOP) transient were calculated. The results show that SARAX code system can give reasonable results of parameter prediction in transient analysis of fast reactor. ULOF transient is more severe for sodium-cooled fast reactor in that it will cause the boiling of sodium and then cause severe accident.
The transient analysis of unprotected accident is significant in safety analysis of sodium-cooled fast reactor. Based on MOX-3600 and MET-1000 benchmarks published by OECD/NEA, SARAX code system was applied to do transient calculation for different sodium-cooled fast reactors. Various reactivity feedback effects in the reactor were analyzed, and the changes of fuel temperature and coolant temperature during unprotected loss of flow(ULOF) transient and unprotected transient over power(UTOP) transient were calculated. The results show that SARAX code system can give reasonable results of parameter prediction in transient analysis of fast reactor. ULOF transient is more severe for sodium-cooled fast reactor in that it will cause the boiling of sodium and then cause severe accident.
引文
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[3] 郑友琦,吴宏春,杜夏楠,等.快堆中子学计算程序NECP-SARAX的改进与验证[J].强激光与粒子束,2017,29(5):29056005.ZHENG Youqi,WU Hongchun,DU Xia’nan,et al.Improvement and verification of the fast reactor neutronics code NECP-SARAX[J].High Power Laser and Particle Beams,2017,29(5):29056005(in Chinese).
[4] ZHENG Youqi,QIAO Liang,ZHAI Zi’an,et al.SARAX:A new code for fast reactor analysis,Part Ⅱ:Verification,validation and uncertainty quantification[J].Nuclear Engineering and Design,2018,331:41-53.
[5] 陈文振,张帆,黎浩峰.核动力点堆中子动力学研究进展[C]//反应堆数值计算与粒子输运学术会议暨2010年反应堆物理会议.[出版地不详]:[出版者不详],2010.
[6] HE Mingtao,WU Hongchun,ZHENG Youqi,et al.Beam transient analyses of accelerator driven subcritical reactors based on neutron transport method[J].Nuclear Engineering and Design,2015,295:489-499.
[7] DU X,ZHENG Y,CAO L,et al.Transient analysis of MOX-3600 and MET-1000 sodium-cooled fast reactor using SARAX code system[J].Annals of Nuclear Energy,2018,121:324-334.
[8] 谢仲生,邓力.中子输运理论数值计算方法[M].西安:西北工业大学出版社,2005.
[9] 徐李,马大园,施工,等.快堆三维瞬态反应性反馈计算研究[J].原子能科学技术,2013,47(10):1 700-1 706.XU Li,MA Dayuan,SHI Gong,et al.Research of three-dimensional transient reactivity feedback in fast reactor[J].Atomic Energy Science and Technology,2013,47(10):1 700-1 706(in Chinese).
[10] BUISON L,STAUFF N.AEN-WPRS benchmark for uncertainty analysis in modeling (UAM) for design,operation and safety analysis of SFRs[R].[S.l.]:OECD/NEA,2017.