Study on IVR Capability Margin of Pressurized Water Reactor Based on Decay Heat Uncertainty
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摘要: 为确定衰变热对高功率压水堆熔融物堆内滞留(IVR)能力边际的影响,采用显著性水平评价与抽样失效率相结合的评价方式,对IVR能力边际进行评价。利用熔融物堆内滞留分析工具CISER开展抽样计算,获得不同核电厂电功率水平、不同衰变热分布参数条件下的下封头壁面热流密度峰值与当地临界热流密度(CHF)的比值,对热流密度比分别开展显著性水平估算与失效率计算,根据小于局部CHF的下封头熔穿准则,判定IVR措施是否有效,以获得IVR能力边际。研究结果表明,若不对下封头内外传热构成进行任何优化措施,电功率超过1400 MW压水堆电厂不推荐单独使用IVR作为严重事故条件缓解措施。
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关键词:
- 衰变热 /
- 临界热流密度(CHF) /
- 熔融物堆内滞留(IVR) /
- 抽样失效率
Abstract: In order to determine the influence of decay heat on the IVR (In-Vessel Retention) capability margin of HPWR, an evaluation method combining significance level evaluation and sampling failure rate was used to evaluate IVR capability margin. Using CISER for IVR to carry out the sampling calculation, the ratio of the peak heat flux on the lower head wall to the local critical heat flux (CHF) under different power levels and decay heat distribution parameters of nuclear power plants was obtained to carry out the significance level estimation and failure rate calculation of the heat flux ratio to judge whether the IVR measure is effective based on the penetration criterion of lower head (less than local CHF) to obtain the IVR capability margin. The results show that IVR alone is not recommended as a serious accident mitigation measure for PWR plants with an electrical power of more than 1400 MW without any optimization of the heat transfer composition inside and outside the lower head.-
Key words:
- Decay heat /
- Critical heat flux (CHF) /
- In-vessel retention (IVR) /
- Sampling failure rate
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表 1 IVR熔池参数
Table 1. IVR Melt Pool Parameters
表 2 样本容量n=1500时典型CHF评价模型的成功率
Table 2. Success Rate of Representative CHF Evaluation Model for Sampling Size n=1500
额定电功率/MW σc 典型CHF评价模型 UCSB INEEL 多孔介质涂层 强化绝热层 联合强化模型 界面脱离模型 综合CHF模型 1400 0.4 0.9338 1.0 0.981 0.992 0.993 0.996 0.968 0.5 0.8932 0.9758 0.953 0.981 0.983 0.985 0.935 0.6 0.8484 0.9226 0.922 0.96 0.969 0.97 0.906 1500 0.4 0.894 0.9975 0.971 0.988 0.991 0.993 0.954 0.5 0.876 0.9465 0.939 0.975 0.981 0.981 0.918 0.6 0.8565 0.909 0.913 0.951 0.958 0.962 0.888 1600 0.4 0.8528 0.9552 0.959 0.985 0.988 0.988 0.938 0.5 0.8448 0.9408 0.923 0.968 0.974 0.975 0.904 0.6 0.8384 0.9136 0.898 0.94 0.951 0.953 0.864 1700 0.4 0.8109 0.9146 0.948 0.981 0.985 0.985 0.919 0.5 0.8245 0.9044 0.914 0.956 0.968 0.969 0.883 0.6 0.8347 0.8993 0.881 0.929 0.939 0.946 0.835 -
[1] REMPE J L, SUH K Y, CHEUNG F B, et al. In-Vessel retention strategy for high power reactors: INEEL/EXT-04-02561[R]. Idaho Falls: Idaho National Engineering and Environmental Laboratory, 2005. [2] THEOFANOUS T G, LIU C, ADDITON S, et al. In-vessel coolability and retention of a core melt[J]. Nuclear Engineering and Design, 1997, 169(1-3): 1-48. doi: 10.1016/S0029-5493(97)00009-5 [3] 付霄华. SCDAP/RELAP5与MELCOR程序对堆芯损伤过程预测的比较[J]. 核动力工程,2003, 24(5): 430-434. doi: 10.3969/j.issn.0258-0926.2003.05.008 [4] REMPE J L, KNUDSON D L, ALLISON C M, et al. Potential for AP600 in-vessel retention through ex-vessel flooding: INEEL/EXT-97-00779[R]. Idaho: Idaho National Engineering and Environmental Laboratory, 1997. [5] 向清安,关仲华,邓纯锐,等. AP1000 IVR三层熔融池结构评价分析[J]. 核动力工程,2013, 34(6): 83-87. doi: 10.3969/j.issn.0258-0926.2013.06.020 [6] ZHANG Y P, QIU S Z, SU G H, et al. Analysis of safety margin of in-vessel retention for AP1000[J]. Nuclear Engineering and Design, 2010, 240(8): 2023-2033. doi: 10.1016/j.nucengdes.2010.04.020 [7] 陈彬, 邹志强. CISER程序用户手册[R]. 中国核动力研究设计院科技报告, 2012 [8] CHEUNG F B, HADDAD K H, LIU Y C. Critical heat flux (CHF) phenomenon on a downward facing curved surface: NUREG/CR-6507[R]. Washington: U.S. Nuclear Regulatory Commission, 1997. [9] ESMAILI H, KHATIB-RAHBAR M. Analysis of likelihood of lower head failure and ex-vessel fuel coolant interaction energetics for AP1000[J]. Nuclear Engineering and Design, 2005, 235(15): 1583-1605. doi: 10.1016/j.nucengdes.2005.02.003 [10] 余红星,苏光辉,关仲华,等. 严重事故条件下反应堆压力容器下封头CHF模型应用研究[J]. 核动力工程,2012, 33(6): 46-50. doi: 10.3969/j.issn.0258-0926.2012.06.011 [11] 鲍晗,金越,刘晓晶,等. 大功率先进压水堆IVR有效性评价中熔池换热研究[J]. 原子能科学技术,2014, 48(2): 234-240. doi: 10.7538/yzk.2014.48.02.0234 [12] 文青龙,陈军,赵华. 倾斜限制空间内池式沸腾临界热流密度分析模型研究[J]. 核动力工程,2010, 31(S1): 109-113+118. [13] 曹克美,许以全,史国宝,等. 严重事故下反应堆压力容器外水冷有效性概率分析[J]. 核动力工程,2009, 30(1): 1-4. [14] REMPE J L, KNUDSON D L. Margin for in-vessel retention in the APR1400 VESTA and SCDAP/RELAP5-3D analyses: INEEL/EXT-04-02549[R]. Idaho Falls: Idaho National Laboratory, 2014. [15] 李治刚,安萍,明平洲,等. 基于方差分解的堆芯下封头熔池模型敏感性分析[J]. 原子能科学技术,2020, 54(8): 1409-1417. doi: 10.7538/yzk.2019.youxian.0501