压水堆核电站大破口失水事故分析
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摘要
压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件,大破口失水事故是核电站运行的设计基准事故,是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算,对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值(PCT)最高,且长时间维持在较高温度,此条件下反应堆最危险。计算结果表明,事故发生后,一回路压力迅速下降,堆芯冷却剂的流动性变差,导致堆芯裸露,燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作,能保证燃料元件包壳温度不超过1 204℃的限值。
The safety analysis report of the pressurized water reactor(PWR) nuclear power plant is an important document for the safety review of the nuclear safety supervision department. The large break loss of coolant accident is a design basis accident for the operation of nuclear power plants and an important part of the safety analysis report. In this paper, RELAP5/MOD3.2 was used to calculate the large break loss of coolant accident of the PWR cold-leg section. It is found that the peak of fuel element cladding temperature(PCT) is the highest when the double-end fracture occurs in the cold-leg section of the primary loop and maintains at a higher temperature for a long time, so the reactor is the most dangerous in this condition. The calculation results show that the pressure of the primary loop drops rapidly after the double-end fracture accident, and the fluidity of the core coolant deteriorates, resulting in the core exposed and the fuel cladding temperature rising again. Through a series of actions such as an injection system and an auxiliary water supply system, it is possible to ensure that the fuel element cladding temperature does not exceed the limit of 1 204 ℃.
The safety analysis report of the pressurized water reactor(PWR) nuclear power plant is an important document for the safety review of the nuclear safety supervision department. The large break loss of coolant accident is a design basis accident for the operation of nuclear power plants and an important part of the safety analysis report. In this paper, RELAP5/MOD3.2 was used to calculate the large break loss of coolant accident of the PWR cold-leg section. It is found that the peak of fuel element cladding temperature(PCT) is the highest when the double-end fracture occurs in the cold-leg section of the primary loop and maintains at a higher temperature for a long time, so the reactor is the most dangerous in this condition. The calculation results show that the pressure of the primary loop drops rapidly after the double-end fracture accident, and the fluidity of the core coolant deteriorates, resulting in the core exposed and the fuel cladding temperature rising again. Through a series of actions such as an injection system and an auxiliary water supply system, it is possible to ensure that the fuel element cladding temperature does not exceed the limit of 1 204 ℃.
引文
[1] USNRC CFR 50—1974 Chapter 50.46:Acceptance criteria for emergency core cooling systems for light water nuclear power reactors—Appendix K to Part 50:ECCS evaluation models[S].US:NRC,1974.
[2] 倪超,匡波,任志豪,等.基于先进程序+保守评价模型的300 MW压水堆核电站大破口失水事故分析[J].原子能科学技术,2012,46(3):328-335.NI Chao,KUANG Bo,REN Zhihao,et al.300 MW PWR NPP LBLOCA analysis based on advanced code plus conservative evaluation models[J].Atomic Energy Science and Technology,2012,46(3):328-335(in Chinese).
[3] 张龙飞,张大发,徐金良.压水堆大破口失水事故高压安注的缓解能力研究[J].核动力工程,2008,29(4):108-111.ZHANG Longfei,ZHANG Dafa,XU Jinliang.Study on mitigating capability of high-pressure safety injection for large break LOCA in PWR[J].Nuclear Power Engineering,2008,29(4):108-111(in Chinese).
[4] OKUBO T,MURAO Y.Assessment of core thermo-hydrodynamic models of REFLA-1D code with CCTF data for reflood phase of PWR-LOCA[J].Journal of Nuclear Science and Technology,1983,22(12):983-994.
[5] IGUCHI T,IWAMURA T,AKIMOTO H.SCTF-Ⅲ test plan and recent SCTF-Ⅲ test results[J].Nuclear Engineering and Design,1988,108(1):241-247.
[6] CHO S,PARK H,CHOI K,et al.Core thermal hydraulic behavior during the reflood phase of cold-leg LBLOCA experiments using the atlas test facility[J].Nuclear Engineering & Technology,2009,41(10):1 263-1 274.
[7] LEE S H,HAN G K.Pre and post test analysis of LBLOCA late reflood phase in ATLAS Using RELAP5/MOD3.3[C]//16th International Conference on Nuclear Engineering.[S.l.]:[s.n.],2008:275-282.
[8] 余建辉,张经瑜,郑利民.大破口LOCA事故ASTRUM最佳估算分析方法优化研究[J].核技术,2014,37(9):78-82.YU Jianhui,ZHANG Jingyu,ZHENG Limin.Study on ASTRUM optimization for large break LOCA analysis[J].Nuclear Techniques,2014,37(9):78-82(in Chinese).
[9] 郭连城,曹学武.核电厂大LOCA始发严重事故下氢气源项的敏感性分析[J].核动力工程,2007,28(5):69-74,108.GUO Liancheng,CAO Xuewu.Sensitivity analysis of hydrogen source during severe accident induced by large LOCA for NPP[J].Nuclear Power Engineering,2007,28(5):69-74,108(in Chinese).
[10] 殷煜皓.AP1000先进核电厂大破口RELAP5建模及特性分析[D].上海:上海交通大学,2012.
[11] 尤伟,石雪垚,王晓霞,等.核电厂大破口失水事故(LBLOCA)始发严重事故释放源项的分析[J].核安全,2015,14(2):53-57.YOU Wei,SHI Xueyao,WANG Xiaoxia,et al.Source term analysis of severe accident induced by LBLOCA for nuclear power plant[J].Nuclear Safety,2015,14(2):53-57(in Chinese).
[12] 骆邦其.冷热段同时安注时的大破口失水事故分析[J].核动力工程,1996,17(5):391-394,406.LUO Bangqi.Analysis of large break loss of coolant accident at cold leg and hot leg simultaneous safety injection[J].Nuclear Power Engineering,1996,17(5):391-394,406(in Chinese).
[13] 骆邦其,孙吉良.CPR1000 核电厂大破口失水事故分析[C]//中国核学会2011 年学术年会论文集第3册(核能动力分卷(下)).北京:中国核学会,2011.
[14] 张龙飞,张大发,王少明.压水堆大破口失水事故引发的严重事故研究[J].原子能科学技术,2007,41(5):560-564.ZHANG Longfei,ZHANG Dafa,WANG Shao-ming.Study on severe accident induced by large break loss of coolant accident for pressurized water reactor[J].Atomic Energy Science and Technology,2007,41(5):560-564(in Chinese).
[15] SATTISON M B,HALL K W.Analysis of core damage frequency:Zion,unit 1 internal events[R].USA:Nuclear Regulatory Commission,1990.
[16] 田文喜,秋穗正,苏光辉,等.CARR热工水力与安全分析程序TSACC的开发与验证[J].核动力工程,2009,30(1):40-44.TIAN Wenxi,QIU Suizheng,SU Guanghui,et al.Development and validation of thermal-hydraulic and safety analysis code for CARR (TSACC)[J].Nuclear Power Engineering,2009,30(1):40-44(in Chinese).
[17] SLOAN S M,SCHULTZ R R,WILSON G E.RELAP5/MOD3 code manual[R].USA:Idaho National Engineering Laboratory,1994.
[18] TODREAS N E,KAZIMI M S.Nuclear reactor thermal analysis[M].USA:[s.n.],1988.
[19] 骆邦其.中破口失水事故的峰值包壳温度与破口等效直径[C]//中国核学会2011年学术年会论文集第3册(核能动力分卷(下)).北京:中国核学会,2011.
[2] 倪超,匡波,任志豪,等.基于先进程序+保守评价模型的300 MW压水堆核电站大破口失水事故分析[J].原子能科学技术,2012,46(3):328-335.NI Chao,KUANG Bo,REN Zhihao,et al.300 MW PWR NPP LBLOCA analysis based on advanced code plus conservative evaluation models[J].Atomic Energy Science and Technology,2012,46(3):328-335(in Chinese).
[3] 张龙飞,张大发,徐金良.压水堆大破口失水事故高压安注的缓解能力研究[J].核动力工程,2008,29(4):108-111.ZHANG Longfei,ZHANG Dafa,XU Jinliang.Study on mitigating capability of high-pressure safety injection for large break LOCA in PWR[J].Nuclear Power Engineering,2008,29(4):108-111(in Chinese).
[4] OKUBO T,MURAO Y.Assessment of core thermo-hydrodynamic models of REFLA-1D code with CCTF data for reflood phase of PWR-LOCA[J].Journal of Nuclear Science and Technology,1983,22(12):983-994.
[5] IGUCHI T,IWAMURA T,AKIMOTO H.SCTF-Ⅲ test plan and recent SCTF-Ⅲ test results[J].Nuclear Engineering and Design,1988,108(1):241-247.
[6] CHO S,PARK H,CHOI K,et al.Core thermal hydraulic behavior during the reflood phase of cold-leg LBLOCA experiments using the atlas test facility[J].Nuclear Engineering & Technology,2009,41(10):1 263-1 274.
[7] LEE S H,HAN G K.Pre and post test analysis of LBLOCA late reflood phase in ATLAS Using RELAP5/MOD3.3[C]//16th International Conference on Nuclear Engineering.[S.l.]:[s.n.],2008:275-282.
[8] 余建辉,张经瑜,郑利民.大破口LOCA事故ASTRUM最佳估算分析方法优化研究[J].核技术,2014,37(9):78-82.YU Jianhui,ZHANG Jingyu,ZHENG Limin.Study on ASTRUM optimization for large break LOCA analysis[J].Nuclear Techniques,2014,37(9):78-82(in Chinese).
[9] 郭连城,曹学武.核电厂大LOCA始发严重事故下氢气源项的敏感性分析[J].核动力工程,2007,28(5):69-74,108.GUO Liancheng,CAO Xuewu.Sensitivity analysis of hydrogen source during severe accident induced by large LOCA for NPP[J].Nuclear Power Engineering,2007,28(5):69-74,108(in Chinese).
[10] 殷煜皓.AP1000先进核电厂大破口RELAP5建模及特性分析[D].上海:上海交通大学,2012.
[11] 尤伟,石雪垚,王晓霞,等.核电厂大破口失水事故(LBLOCA)始发严重事故释放源项的分析[J].核安全,2015,14(2):53-57.YOU Wei,SHI Xueyao,WANG Xiaoxia,et al.Source term analysis of severe accident induced by LBLOCA for nuclear power plant[J].Nuclear Safety,2015,14(2):53-57(in Chinese).
[12] 骆邦其.冷热段同时安注时的大破口失水事故分析[J].核动力工程,1996,17(5):391-394,406.LUO Bangqi.Analysis of large break loss of coolant accident at cold leg and hot leg simultaneous safety injection[J].Nuclear Power Engineering,1996,17(5):391-394,406(in Chinese).
[13] 骆邦其,孙吉良.CPR1000 核电厂大破口失水事故分析[C]//中国核学会2011 年学术年会论文集第3册(核能动力分卷(下)).北京:中国核学会,2011.
[14] 张龙飞,张大发,王少明.压水堆大破口失水事故引发的严重事故研究[J].原子能科学技术,2007,41(5):560-564.ZHANG Longfei,ZHANG Dafa,WANG Shao-ming.Study on severe accident induced by large break loss of coolant accident for pressurized water reactor[J].Atomic Energy Science and Technology,2007,41(5):560-564(in Chinese).
[15] SATTISON M B,HALL K W.Analysis of core damage frequency:Zion,unit 1 internal events[R].USA:Nuclear Regulatory Commission,1990.
[16] 田文喜,秋穗正,苏光辉,等.CARR热工水力与安全分析程序TSACC的开发与验证[J].核动力工程,2009,30(1):40-44.TIAN Wenxi,QIU Suizheng,SU Guanghui,et al.Development and validation of thermal-hydraulic and safety analysis code for CARR (TSACC)[J].Nuclear Power Engineering,2009,30(1):40-44(in Chinese).
[17] SLOAN S M,SCHULTZ R R,WILSON G E.RELAP5/MOD3 code manual[R].USA:Idaho National Engineering Laboratory,1994.
[18] TODREAS N E,KAZIMI M S.Nuclear reactor thermal analysis[M].USA:[s.n.],1988.
[19] 骆邦其.中破口失水事故的峰值包壳温度与破口等效直径[C]//中国核学会2011年学术年会论文集第3册(核能动力分卷(下)).北京:中国核学会,2011.