Aging Evaluation of Baffle Bolts for Reactor Internals of Qinshan Nuclear Power Plant
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摘要: 介绍了秦山核电厂为评估堆内构件围板螺栓的实际老化状态，在吸收总结国际上堆内构件老化机理研究成果的基础上建立了堆内构件围板螺栓的老化机理判断准则，用其评估识别围板螺栓主要受磨损、应力松弛、辐照肿胀、辐照促进应力腐蚀开裂等老化机理的影响，并针对老化机理可能导致的缺陷类型，开发了水下超声检查技术补充常规的目视检查方法，从而制定评估老化状态的检查方案。评估结果表明，秦山核电厂堆内构件围板螺栓老化状态良好，尚未发生变形、裂纹等老化失效现象。实践证明该评估方法行之有效，可用于压水堆核电厂部件老化评估工作。Abstract: In order to evaluate the actual aging state of baffle bolts for reactor internals in Qinshan Nuclear Power Plant, on the basis of absorbing and summarizing the international research results on the aging mechanism of reactor internals, a criterion for judging the aging mechanism of reactor internals is established, and is used to evaluate and identify the aging mechanism of baffle bolts mainly affected by wear, stress relaxation, irradiation swelling, irradiation-promoted stress corrosion cracking and so on. Besides, according to the possible defect types caused by the aging mechanism, the underwater ultrasonic inspection technology is developed to supplement the conventional visual inspection method, so as to formulate the inspection scheme to evaluate the aging state. The evaluation results show that the baffle bolts for reactor internals in Qinshan Nuclear Power Plant are in good aging condition, and no deformation, crack and other aging failure phenomena have occurred. Practice has proved that the evaluation method is effective and can be used for the aging evaluation of PWR nuclear power plant components.
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Key words:
- Reactor internals /
- Baffle bolts /
- Aging evaluation /
- UT inspection
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图 1 应力腐蚀开裂机理分析评估
Figure 1. Analysis and Evaluation of Stress Corrosion Cracking Mechanism
图 3 满功率运行时围板-辐板-吊篮局部应力强度分布
Figure 3. Local Stress Intensity Distribution of Baffle-Former-Basket under Full Power Operation
表 1 围板螺栓老化机理
Table 1. Aging Mechanism of Baffle Bolts
老化机理 材料 筛选准则 磨损 所有合金 可能发生相对运动的相邻部件；
或者要求有预紧力的位置；
或者发生应力松弛/辐照蠕变的位置应力松弛 所有材料 辐照增强的应力松弛和蠕变，中子注量≥0.2 dpa [1.3×1020 cm−2 (E>1.0 MeV)]，适用于所有螺栓或弹簧位置 辐照肿胀 奥氏体不锈钢或其他材料 中子注量≥20 dpa [1.3×1022 cm−2 (E>1.0MeV)]；温度≥320℃ 辐照脆化 奥氏体不锈钢 中子注量≥1.5 dpa [1.0×1021 cm−2 (E>1.0 MeV)] 应力腐蚀开裂 镍基合金 筛选应力≥207 MPa 辐照促进应力腐蚀开裂（IASCC） 所有材料 中子注量<3 dpa [2.0×1021 cm−2 (E>1.0 MeV)]，参考应力腐蚀开裂筛选准则 高中子注量＞3 dpa [2.0×1021 cm−2 (E>1.0 MeV)]），筛选应力= −158.4 ln(dpa) + 789.65（MPa） 疲劳 所有合金 U≥0.1或适用于筛选出应力松弛或辐照蠕变的位置 E—弹性模量 
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表 2 围板和辐板的中子注量
Table 2. Neutron Fluence in Baffle and Former
部件名称 中子注量/（1021 cm−2） (σ/σo)/% 围板 38.30 0 幅板 2.34~38.30 0~36.2 螺栓 38.30 0
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表 3 部件低周疲劳分析结果
Table 3. Low Cycle Fatigue Analysis Results of Components
部件 U 围板 7.09×10－5 辐板 1.03×10－7 吊篮 7.04×10－5
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[1] 中核核电运行管理有限公司. 堆内构件老化管理大纲[Z]. 2018. [2] 中核核电运行管理有限公司. 秦山320MWe机组运行许可证延续项目堆内构件老化管理审查报告[Z]. 2021. [3] 中核核电运行管理有限公司. 秦一厂118大修反应堆压力容器堆内构件电视检验结果综合报告[Z]. 2018. [4] 中核核电运行管理有限公司. 秦一厂118大修反应堆压力容器堆内构件围板螺栓超声检查综合报告[Z]. 2018. -
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