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[1]黄志翱,张泽枫,缪惠芳*,等.严重事故下蒸汽发生器传热管诱发破裂现象及其缓解策略分析[J].厦门大学学报(自然科学版),2019,58(02):260-268.[doi:10.6043/j.issn.0438-0479.201809022]
 HUANG Zhiao,ZHANG Zefeng,MIAO Huifang*,et al.Analysis of induced steam generator tube rupture phenomena and its mitigation strategies under severe accidents[J].Journal of Xiamen University(Natural Science),2019,58(02):260-268.[doi:10.6043/j.issn.0438-0479.201809022]
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严重事故下蒸汽发生器传热管诱发破裂现象及其缓解策略分析(PDF/HTML)
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《厦门大学学报(自然科学版)》[ISSN:0438-0479/CN:35-1070/N]

卷:
58卷
期数:
2019年02期
页码:
260-268
栏目:
研究论文
出版日期:
2019-03-27

文章信息/Info

Title:
Analysis of induced steam generator tube rupture phenomena and its mitigation strategies under severe accidents
文章编号:
0438-0479(2019)02-0260-09
作者:
黄志翱1张泽枫2 缪惠芳1*李 宁1
1.厦门大学能源学院,福建 厦门 361102; 2.苏州热工研究院有限公司深圳分公司,广东 深圳 518000
Author(s):
HUANG Zhiao1ZHANG Zefeng2MIAO Huifang1*LI Ning1
1.College of Energy,Xiamen University,Xiamen 361102,China; 2.Suzhou Thermal Engineering Research Institute Co.,Ltd.Shenzhen Branch,Shenzhen 518000,China
关键词:
蒸汽发生器传热管 诱发破裂 逆向自然循环 卸压 补水 CPR1000
Keywords:
steam generator tube induced rupture counter-current natural circulation bleed feed CPR1000
分类号:
TM 623
DOI:
10.6043/j.issn.0438-0479.201809022
文献标志码:
A
摘要:
蒸汽发生器传热管是核反应堆冷却剂系统压力边界的重要组成部分,研究严重事故下蒸汽发生器传热管诱发破裂现象及其影响因素对支持二级概率安全分析意义重大.以CPR1000电厂全厂断电叠加蒸汽发生器安全阀卡开事故为基础事故序列,分析了轴封破口、环路水封清除和下降管水封清除现象对蒸汽发生器传热管诱发蠕变破裂现象的影响,并对二次侧卸压-补水和一次侧卸压-补水两种缓解策略的效果进行了研究.结果表明:轴封破口现象会影响逆向自然循环流量,但不会影响热管段和蒸汽发生器传热管发生蠕变破裂的先后顺序; 而环路水封清除和下降管水封清除现象会打破热管段逆向自然循环现象,并导致蒸汽发生器传热管比其他冷却剂系统边界更早失效,从而带来安全壳旁通风险; 而二次侧卸压-补水策略和一次侧卸压-补水策略都可以达到降低蒸汽发生器传热管诱发破裂风险的效果.该研究结果有助于改进二级概率安全分析结果,指导CPR1000电厂制定相关严重事故缓解措施并提升严重事故管理导则的事故处置能力.
Abstract:
Steam generator(SG)tubes are a substantial portion of the reactor coolant pressure boundary(RCPB).Analysis of severe accident induced steam generator tube rupture(SAI-SGTR)phenomena is of great importance for level 2 probabilistic safety assessment(PSA).The base case in this paper is steam generator safety valve stuck-open accident combined with station blackout(SBO)in the CPR1000 power plant.In addition,the influence of seal loss-of-coolant accident(LOCA),loop seal clear and downcomer seal clear phenomena on the SAI-SGTR results were analyzed based on the base case.Our analyses indicate that the occurrence of seal LOCA has influence on the flow rate of the hot leg counter-current natural circulation,but it cannot change the sequence of the occurrence of hot leg creep rupture(HLCR)and SAI-SGTR.However,it is observed that the loop seal clear and downcomer seal clear phenomena can break the original hot leg counter-current natural circulation and lead to earlier occurrence of SAI-SGTR than that of other RCPB,which results in the containment bypass risk in the end.Moreover,both the secondary bleed-and-feed strategy and the primary bleed-and-feed strategy show great mitigation effectiveness to lower the risk of induced-SGTR.The results of this study are helpful to improve the results of level 2 PSA,to guide the CPR1000 power plant to develop relevant severe accident mitigation strategies,and to enhance the accident handling ability of severe accident management guidelines(SAMGs).Steam generator(SG)tubes are a substantial portion of the reactor coolant pressure boundary(RCPB).Analysis of severe accident induced steam generator tube rupture(SAI-SGTR)phenomena is of great importance for level 2 probabilistic safety assessment(PSA).The base case in this paper is steam generator safety valve stuck-open accident combined with station blackout(SBO)in the CPR1000 power plant.In addition,the influence of seal loss-of-coolant accident(LOCA),loop seal clear and downcomer seal clear phenomena on the SAI-SGTR results were analyzed based on the base case.Our analyses indicate that the occurrence of seal LOCA has influence on the flow rate of the hot leg counter-current natural circulation,but it cannot change the sequence of the occurrence of hot leg creep rupture(HLCR)and SAI-SGTR.However,it is observed that the loop seal clear and downcomer seal clear phenomena can break the original hot leg counter-current natural circulation and lead to earlier occurrence of SAI-SGTR than that of other RCPB,which results in the containment bypass risk in the end.Moreover,both the secondary bleed-and-feed strategy and the primary bleed-and-feed strategy show great mitigation effectiveness to lower the risk of induced-SGTR.The results of this study are helpful to improve the results of level 2 PSA,to guide the CPR1000 power plant to develop relevant severe accident mitigation strategies,and to enhance the accident handling ability of severe accident management guidelines(SAMGs).

参考文献/References:

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[2] 胡文超,彭常宏.严重事故时蒸汽发生器传热管蠕变断裂风险评估[J].核科学与技术,2015,3(3):49-54.
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[7] ADER C,COLLINS T,DONOGHUE J,et al.Risk assessment of severe accident induced steam generator tube rupture,NUREG-1570[R].North Bethesda:Nuclear Regulatory Commission,1998.
[8] SARBES A,JAMES G,BHARAT A,et al.Severe accident risks:an assessment for five U.S.nuclear power plants,NUREG-1150[R].North Bethesda:Nuclear Regulatory Commission,1990.
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[10] FULLER E L,KENTON M A,EPSTEIN M,et al.Steam generator tube integrity risk assessment volume 1:general methodology,EPRI TR-107623[R].Palo Alto:Electric Power Research Institute,2002.
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[13] DENG J,CAO X W.Analysis of hot leg natural circulation under station blackout severe accident[J].Nuclear Science and Techniques,2007,18(2):123-128.
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[15] 杨奥,黄志翱,缪惠芳,等.CPR1000全厂断电事故模拟及主泵轴封破口敏感性分析[J].厦门大学学报(自然科学版),2018,57(5):629-633.
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[2] 胡文超,彭常宏.严重事故时蒸汽发生器传热管蠕变断裂风险评估[J].核科学与技术,2015,3(3):49-54.
[3] 陈宝文,毛欢,孔翔程,等.全厂断电引发的严重事故下蒸汽发生器传热管蠕变失效风险研究[J].原子能科学与技术,2014,48(6):1026-1030.
[4] BAYLESS P D,BROWNSON D A,DOBBE C A,et al.Severe accident natural circulation studies at the INEL,NUREG/CR-6285[R].North Bethesda:Nuclear Regulatory Commission,1995.
[5] MACDONALD P E,SHAH V N,WARD L W,et al.Steam generator tube failures,NUREG/CR-6365[R].North Bethesda:Nuclear Regulatory Commission,1996.
[6] DIERCKS D R,BAKHTIARI S,KASZA K E,et al.Steam generator tube integrity program,NUREG/CR-6511[R].North Bethesda:Nuclear Regulatory Commission,1997.
[7] ADER C,COLLINS T,DONOGHUE J,et al.Risk assessment of severe accident induced steam generator tube rupture,NUREG-1570[R].North Bethesda:Nuclear Regulatory Commission,1998.
[8] SARBES A,JAMES G,BHARAT A,et al.Severe accident risks:an assessment for five U.S.nuclear power plants,NUREG-1150[R].North Bethesda:Nuclear Regulatory Commission,1990.
[9] FLETCHER C D,BEATON R M,PALAZOV V V,et al.SCDAP/RELAP5 thermal-hydraulic evaluations of the potential for containment bypass during extended station blackout severe accident sequences in a westinghouse four-loop PWR,NUREG/CR-6995[R].North Bethesda:Nuclear Regulatory Commission,2009.
[10] FULLER E L,KENTON M A,EPSTEIN M,et al.Steam generator tube integrity risk assessment volume 1:general methodology,EPRI TR-107623[R].Palo Alto:Electric Power Research Institute,2002.
[11] FULLER E L,KENTON M A,EPSTEIN M,et al.Steam generator tube integrity risk assessment volume 2:application to Diablo Canyon Power,EPRI TR-107623[R].Palo Alto:Electric Power Research Institute,2006.
[12] ELECTRIC POWER RESEARCH INSTITUTE.MAAP4 applications guidance:desktop reference for using MAAP4 software,EPRI TR-1020236[R].Palo Alto:Electric Power Research Institute,2005.
[13] DENG J,CAO X W.Analysis of hot leg natural circulation under station blackout severe accident[J].Nuclear Science and Techniques,2007,18(2):123-128.
[14] LIAO Y,VIEROW K.MELCOR analysis of steam generator tube creep rupture in station blackout severe accident[J].Nuclear Technology,2005,545(1):30-35.
[15] 杨奥,黄志翱,缪惠芳,等.CPR1000全厂断电事故模拟及主泵轴封破口敏感性分析[J].厦门大学学报(自然科学版),2018,57(5):629-633.
[16] MCHUGH P R,HENTZEN R D.Natural circulation cooling in U.S.pressurized water reactors,NUREG/CR-5769[R].North Bethesda:Nuclear Regulatory Commission,1992.

备注/Memo

备注/Memo:
收稿日期:2018-09-16 录用日期:2018-12-05
基金项目:福建省科技厅计划项目(2016H0034); 厦门大学能源学院发展基金(2017NYFZ01)
*通信作者:hfmiao@xmu.edu.cn
更新日期/Last Update: 1900-01-01