[1]郭裕,姚世卫,李邦明,等.压水堆核电厂管道贯穿裂纹泄漏率计算分析[J].应用科技,2018,45(05):108-113.[doi:10.11991/yykj.201712012]
 GUO Yu,YAO Shiwei,LI Bangming,et al.Calculation of the leak rate of through-wall crack in pressurized water reactor nuclear power plant[J].yykj,2018,45(05):108-113.[doi:10.11991/yykj.201712012]
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《应用科技》[ISSN:1009-671X/CN:23-1191/U]

卷:
第45卷
期数:
2018年05期
页码:
108-113
栏目:
核科学技术与应用
出版日期:
2018-09-15

文章信息/Info

Title:
Calculation of the leak rate of through-wall crack in pressurized water reactor nuclear power plant
作者:
郭裕 姚世卫 李邦明 李少丹 李勇
武汉第二船舶设计研究所 热能动力技术重点实验室, 湖北 武汉 430205
Author(s):
GUO Yu YAO Shiwei LI Bangming LI Shaodan LI Yong
Science and Technology on Thermal Energy and Power Lab. Wuhan Second Ship Design and Research Institute, Wuhan 430205, China
关键词:
压水堆贯穿裂纹破前漏泄漏率计算裂纹形貌裂纹张开位移粗糙度临界流动
Keywords:
pressurized water reactorthrough-wall crackleak before breakleak rate calculationcrack morphologycrack opening displacementroughnesscritical flow
分类号:
TL421.1
DOI:
10.11991/yykj.201712012
文献标志码:
A
摘要:
针对压水堆核电厂管道破前漏分析技术,开展贯穿裂纹泄漏率计算分析研究。通过数值仿真确定裂纹流道流动特性,研究裂纹几何形貌对裂纹泄漏率的影响;基于Henry-Fauske临界流理论编制用于压水堆核电厂管道贯穿裂纹泄漏率的计算程序,并将计算结果与实验数据进行对比,结果表明程序预测误差在30%以内。基于所形成的泄漏率计算程序,研究不同参数对泄漏率的影响,对于压水堆核电厂管道贯穿裂纹泄漏率分析具有指导意义。
Abstract:
Based on the analysis of leak before break of pressurized water reactor pipeline in nuclear power plant, the through-wall crack leak rate was calculated and analyzed. The flow characteristics of crack channel were analyzed by numerical simulation, and the influence of crack morphology on the leakage rate of crack was studied. A calculation procedure for the crack leakage rate in pressurized water reactor pipeline was established based on the Henry-Fauske critical flow theory. The calculation results were compared with the experimental data, showing that the computation error of the program was within 30%. The influence of different parameters on leakage rate was studied on the basis of the leak rate calculation program, which are instructive on leak rate of through-wall crack in pressurized water reactor.

参考文献/References:

[1] IAEA. Assessment and management of ageing of major nuclear power plant components important to safety primary piping in PWRs[R]. Vienna:IAEA, 2003.
[2] 周胜, 张征明. 破前漏分析中泄漏率模型研究进展[J]. 原子能科学技术, 2009, 43(S1):84-91
[3] WANG Mingjun, QIU Suizheng, SU Guanghui, et al. Research on the leak-rate characteristics of leak-before-break (LBB) in pressurized water reactor (PWR)[J]. Applied thermal engineering, 2014, 62(1):133-140.
[4] PAUL D D, AHMAD J, SCOTT P M, et al. Evaluation and refinement of leak-rate estimation models. NUREG/CR-5128[R]. Washington:Nuclear Regulatory Commission, 1991.
[5] 吴万军, 谢海, 兰彬, 等. 管道裂纹泄漏率计算软件开发[J]. 核动力工程, 2015, 36(4):65-68
[6] 殷海峰, 梁兵兵, 徐宁. 管道泄漏率计算模型研究和程序开发[J]. 核技术, 2013, 36(4):040618
[7] 乔红威, 李琦, 刘志伟, 等. LBB设计中管道贯穿裂纹张开位移及泄漏率计算研究[J]. 核技术, 2013, 36(4):040619
[8] 章静, 乔红威, 李朋洲, 等. 管道贯穿裂纹泄漏率预测[J]. 原子能科学技术, 2015, 49(4):660-666
[9] PARK J H, CHO Y K, KIM S H, et al. Estimation of leak rate through circumferential cracks in pipes in nuclear power plants[J]. Nuclear engineering and technology, 2015, 47(3):332-339.
[10] HENRY R E. Critical discharge of initially saturated or subcooled liquid[R]. Cleveland:Lewis Research Center, 1969.
[11] HENRY R E, FAUSKE H K, MCCOMAS S T. Two-phase critical flow at low qualities. Part I. experimental[R]. Argonne:Argonne National Laboratory, 1970.
[12] AMOS C N, SCHROCK V E. Critical discharge of initially sub-cooled water through slits. NUREG/CR-3475[R]. Berkeley:Lawrence Berkeley Lab, 1983.
[13] COLLIER R P, STULEN M E, MAYFIELD D B, et al. Two-phase flow through intergranular stress corrosion cracks and resulting acoustic emission. NP-3540-LD[R]. Battelle, Columbus, USA:EPRI Report, 1984.

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备注/Memo

备注/Memo:
收稿日期:2017-12-22。
作者简介:郭裕(1993-),男,硕士研究生;姚世卫(1971-),男,研究员
通讯作者:郭裕,E-mail:guoyu0115@163.com
更新日期/Last Update: 2018-09-04