Irradiation Embrittlement Time-limited Aging Analysis of Reactor Pressure Vessel for Qinshan Nuclear Power Plant
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摘要: 反应堆压力容器是核电厂最重要设备之一，其辐照脆化状态决定了核电厂的实际运行寿命。通过借鉴国外反应堆压力容器安全评估方法，开发出一套反应堆压力容器辐照脆化时限老化分析（TLAA）的方法。该方法从上平台能量、反应堆运行压力-温度曲线及承压热冲击3个方面评价压力容器材料在正常工况和事故工况下的安全裕度。采用该方法在秦山核电厂运行许可证延续（OLE）项目中对反应堆压力容器进行了辐照脆化TLAA安全评估，其评估方法和评估结论到得国家核安全监管局的认可，为秦山核电厂延寿20 a奠定了基础。
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关键词:
- 压水堆 /
- 辐照脆化 /
- 时限老化分析（TLAA） /
- 运行许可证延续（OLE）
Abstract: Reactor pressure vessel is one of the most important equipment in nuclear power plant, and its irradiation embrittlement state determines the actual operation life of nuclear power plant. A method of irradiation embrittlement time-limited aging analysis (TLAA) for reactor pressure vessel is developed by referring to foreign reactor pressure vessel safety assessment methods. This method evaluates the safety margin of pressure vessel materials under normal and accident conditions from three aspects: upper platform energy, reactor operating pressure-temperature curve and pressure bearing thermal shock. Using this method, the irradiation embrittlement TLAA safety evaluation of the reactor pressure vessel is carried out in the Qinshan Nuclear Power Plant Operation License Extension (OLE) Project. Its evaluation methods and evaluation conclusions have been recognized by the National Nuclear Safety Administration, laying the foundation for the 20-year life extension of Qinshan NPP. -
图 1 辐照前后材料冲击性能变化
ΔRTNDT—无塑性转变温度的增加值；$ {{\Delta }}{U_{{\text{SE}}}} $—上平台能量的增加值
Figure 1. Changes of Impact Properties of Materials Before and after Irradiation
图 2 RPV TLAA示意图
PTS—反应堆压力容器承压热冲击；ASME—美国机械工程师协会锅炉与压力容器规范要求；10CFR 50.61—美国联邦法规关于PTS的规范要求；R.G.1.99—反应堆压力容器材料辐照脆化规范要求
Figure 2. TLAA Diagram of RPV
图 3 辐照监督管材料ΔRTNDT变化趋势图
Figure 3. ΔRTNDT Change Trend of Irradiation Monitoring Tube Material
图 5 50 a寿期末反应堆冷却剂P-T限制曲线
Figure 5.
Reactor Coolant P-T Limit Curve at the End of 50 a Service Life 表 1 堆芯带区材料在50 a寿期末的RTPTS计算结果
Table 1. RTPTS Calculation Results of Core Materials at the End of 50 a Service Life
R.G 1.99 CF/℃ M /℃ ΔRTNDT/℃ ART/℃ 内表面 1/4壁厚 内表面 1/4壁厚 母材 推荐 17.22 26.71 24.92 23.54 31.63 30.26 实验 23.80 21.12 34.43 32.53 35.55 33.65 焊缝 推荐 37.78 36.40 29.63 25.48 46.03 41.88 实验 10.04 24.47 7.87 6.77 12.34 11.24 
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[1] 高永健,贺寅彪,曹明,等. 反应堆压力容器堆芯支承块及附近下封头应力分析和评定[J]. 核技术,2013, 36(4): 040659. [2] U. S. Nuclear Regulatory Commission. 1 CFR Part 54, nuclear energy institute, industry guideline for implementing the requirements, the license renewal rule[S]. Washington DC: Nuclear Regulatory Commission, 2011. [3] U. S. Nuclear Regulatory Commission. Regulatory guide, 1.99, revision 2, radiation embrittlement of reactor vessel materials[S]. Washington DC: Nuclear Regulatory Commission, 1988. [4] The American Society of Mechanical Engineers. Boiler and pressure vessel code, section III Appendix G[S]. New York: ASME, 1989. [5] U. S. Nuclear Regulatory Commission. 10 CFR 50.61, Fracture toughness requirements for protection against pressurized thermal shock events[S]. Washington DC: Nuclear Regulatory Commission, 1984. -
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