某核电厂RPV辐照脆化性能预测与延续运行评估

房永刚; 佟振峰; 初起宝; 曾珍; 慎英才

doi:10.13832/j.jnpe.2023.04.0192

某核电厂RPV辐照脆化性能预测与延续运行评估

doi: 10.13832/j.jnpe.2023.04.0192

1.
生态环境部核与辐射安全中心，北京，100082
2.
华北电力大学，北京，102206

基金项目: 国家重点研发计划项目（2019YFB1900900）

详细信息

作者简介:
房永刚（1981—），男，硕士研究生，现主要从事核设备安全与寿命评价审评工作，E-mail: fangyonggang@chinansc.cn

通讯作者:
初起宝，E-mail: chuqibao@chinansc.cn

中图分类号: TL48
计量
- 文章访问数: 133
- HTML全文浏览量: 41
- PDF下载量: 33
- 被引次数: 0
出版历程
- 收稿日期: 2022-11-21
- 修回日期: 2023-05-08
- 刊出日期: 2023-08-15

Prediction of RPV Irradiation Embrittlement Performance and Life Evaluation of Extended Operation in a Nuclear Power Plant

1.
Nuclear and Radiation Safety Center, MEE, Beijing, 100082, China
2.
North China Electric Power University, Beijing, 102206, China

摘要

摘要: 我国自主设计建造的某核电厂已进入延续运行阶段，作为反应堆核心部分的压力容器辐照脆化性能评价采用了国外的辐照脆化预测模型，但该模型基于的辐照数据不能有效代表我国反应堆压力容器（RPV）材料的辐照脆化性能，尤其是针对延续运行阶段。本文基于国内外RPV辐照脆化预测模型及其开发机理，构建了适用于我国工程应用的自主低Cu RPV辐照脆化预测模型，该模型考虑了稳定基体缺陷和合金元素析出沉淀等辐照脆化关键因素，同时根据国产低Cu RPV材料的辐照脆化数据，开展了自主模型的标准偏差和裕量分析，结果表明模型预测置信度较高。最后依据自主模型评估该核电厂RPV的辐照脆化性能，证明其延续运行至60 等效满功率年（EFPY）具有可行性。
- 压力容器 /
- 辐照脆化 /
- 预测模型 /
- 延续运行
Abstract: A domestic self-designed and built nuclear power plant has entered the stage of extended operation. A foreign prediction model was adopted for the irradiation embrittlement evaluation of the reactor pressure vessel, however, the irradiation data on which the foreign prediction model was based cannot effectively represent the irradiation embrittlement performance of domestic RPV materials, especially for the extended operation stage. Based on the radiation embrittlement prediction models of RPV at home and abroad and their development mechanisms, an independent low-Cu RPV radiation embrittlement prediction model suitable for domestic engineering applications has been formed, which considers the key factors of irradiation embrittlement such as stable matrix defects and alloy element precipitation. According to the irradiation embrittlement data of domestic low-Cu RPV materials, the standard deviation and margin analysis of the independent model were carried out, and the results indicated a high level of confidence in prediction. Based on the model, the irradiation performance of the RPV was evaluated, and the feasibility of extended operation to 60 equivalent full power years (EFPY) was demonstrated.
- Reactor pressure vessel /
- Irradiation embrittlement /
- Prediction model /
- Extended operation

HTML全文

图 1 RPV预测模型与低Cu RPV辐照监督数据比较

Figure 1. Comparison of RPV Prediction Model and Low-Cu RPV Irradiation Surveillance Data

下载: 全尺寸图片幻灯片

图 2 低Cu RPV预测模型与低Cu RPV辐照监督数据比较　　　　　　　

Figure 2. Comparison of Low-Cu RPV Prediction Model and Low-Cu RPV Irradiation Surveillance Data

下载: 全尺寸图片幻灯片

图 3 某核电厂RPV中子注量与运行时间的关系

EFPY—等效满功率年　　

Figure 3. Relationship between RPV Neutron Fluence and Operating Time in a Nuclear Power Plant

下载: 全尺寸图片幻灯片

图 4 母材冲击样品USE与中子注量的关系拟合曲线

Figure 4. Fitting Curve of the Relationship between USE and Neutron Fluence of Base Metal Impact Samples

下载: 全尺寸图片幻灯片

图 5 某核电厂RPV关键性能指标评估图

Figure 5. Evaluation of RPV Key Performance Indicators for a Nuclear Power Plant

下载: 全尺寸图片幻灯片

表 1 不同运行时间RPV内表面及1/4壁厚处的中子注量

Table 1. Neutron Fluence at the Inner Surface and 1/4 Wall Thickness of RPV at Different Operating Periods

运行时间 / EFPY	中子注量/10¹⁹cm⁻²
运行时间 / EFPY	堆芯中平面内表面	上焊缝	堆芯中平面 1/4壁厚	上焊缝 1/4壁厚
40	8.37	0.68	5.54	0.45
50	10.46	0.84	6.92	0.56
60	12.56	1.01	8.30	0.67

下载: 导出CSV

表 2 压力容器内表面$ \Delta {T_{{\text{R,NDT}}}} $和ART的计算值

Table 2. $ \Delta {T_{{\text{R,NDT}}}} $ and ART at RPV Internal Surface

运行时间/EFPY	$ \Delta {T_{{\text{R,NDT}}}} $/℃				ART/℃
运行时间/EFPY	堆芯中平面	上焊缝	堆芯带状区1/4壁厚	上焊缝1/4壁厚	堆芯中平面	上焊缝	堆芯带状区1/4壁厚	上焊缝1/4壁厚
40	66	15	56	12	70	19	60	16
50	72	18	61	14	76	22	65	18
60	77	20	66	15	81	24	70	19

下载: 导出CSV

表 3 不同运行时期RPV的USE

Table 3. USE of RPV at Different Operating Periods

参数名	参数值
运行时间/EFPY	40	50	60
USE/J	211	209	206

下载: 导出CSV

表 4 运行60 EFPY后RPV材料性能

Table 4. RPV Material Properties at the End of 60 EFPY

RPV性能指标	预测计算值	寿期限值	引用规范
1/4壁厚处ART/℃	70	93	RG1.99(2) 第三章
USE/J	206	68	10CFR50
T_R_,_PTS /℃	81	132	10CFR50

下载: 导出CSV

参考文献(12)

[1]	杨文斗. 反应堆材料学[M]. 北京: 原子能出版社, 2006: 328.
[2]	American Society for Testing and Materials. Standard practice for design of surveillance programs for light-water moderated nuclear power reactor vessels: ASTM E185[S]. USA: ASTM Committee, 1982: 1-8.
[3]	U. S. Nuclear Regulatory Commission. Radiation embrittlement of reactor vessel materials, regulatory guide 1.99 revision 2: TASK ME 305-4[R]. Washington: U. S. Nuclear Regulatory Commission, 1988.
[4]	日本原子力规则委员会. 核反应堆压力容器材料监督试验方法: JEAC 4201[S]. 日本: 日本电气协会, 1991:12-58.
[5]	TANON A P, GRANDEMANGE J, HOUSSIN B, et al. French verification of PWR vessel integrity: EPRI-NP--6713[R]. Palo Alto: Research Reports Center, 1990.
[6]	EASON E D, WRIGHT J E, ODETTE G R. Improved embrittlement correlations for reactor pressure vessel steels: NUREG/CR-6551[R]. Washington: U. S. Nuclear Regulatory Commission, 1998.
[7]	SERVER W L, ENGLISH C A, NAIMAN D Q, et al. Charpy embrittlement correlations – status of combined mechanistic and statistical bases for U. S. RPV steels: EPRIMRP-45[R]. Palo Alto: Electric Power Research Institute, 2001.
[8]	EASON E D, ODETTE G R, NANSTAD R K. A Physically based correlation of irradiation-induced transition temperature shifts for RPV steels: ORNL/TM-2006/530[R]. Tennessee: Oak Ridge National Laboratory, 2006
[9]	ERICKSONKIRK M. A review of ∆T30 data for reactor pressure vessel steels obtained at high fluence[J]. J ASTM Intl., 2009: JAI102000-8.
[10]	佟振峰,林虎,宁广胜,等. 低铜合金反应堆压力容器钢辐照脆化预测评估模型[J]. 原子能科学技术,2009, 43(S1): 103-108.
[11]	American Society for Testing and Materials. Standard practice for extrapolating reactor vessel surveillance dosimetry results: ASTM E560[S]. USA: ASTM Committee, 2001: 1-5
[12]	U. S. Nuclear Regulatory Commission. 10CFR Part50 domestic licensing of production and utilization facilities, appendix A[Z]. USA: Nuclear Regulatory Commission, 2021.