秦山核电二期工程反应堆压力容器及其寿命预计

介绍了秦山核电二期工程反应堆压力容器(RPV)的设计思想和背景;说明了RPV产品的基本特征;按照NRC-RG1.99(Rev2)规定给出了快中子(E＞1Mev)辐照损伤计算结果;并对RPV的使用寿命进行了计算,结果表明,在堆芯核设计和燃料管理不作任何优化时,其预计寿命依然能够达到60年.     

反应堆压力容器辐照脆化状态评估

基于大量相似辐照脆化试验测试数据和实际辐照监督测试数据,采用统计分析的方法,选出适合于某核电厂反应堆压力容器（RPV）的辐照脆化评估公式。以该核电厂已经完成的辐照监督管测试数据为输入,对RPV当前的辐照脆化状态进行了评估,并推算、分析了RPV在寿期末的结构完整性;基于辐照脆化计算结果,绘制了各运行阶段RPV的压力-温度限值曲线（P-T曲线）,并给出运行建议。      

秦山第二核电厂1、2号机组压力容器辐照脆化监督

秦山第二核电厂1、2号机组的辐照监督试验全部8个批次已经完成。通过不同快中子注量下辐照监督试样的力学试验,得到压力容器材料的强度、延伸率、冲击韧性等的变化值,以及上平台能量和韧脆转变温度变化值。与未辐照材料性能相比,监督管试样均有一定的辐照强化和脆化效应,韧脆转变温度有不同程度的上升,但均低于FIS公式预测值。综合评价认为:2个机组压力容器材料的辐照脆化效应均处于较低水平。     

国产反应堆压力容器的辐照脆化行为及预测

反应堆压力容器（RPV）是保障核电站运行安全性、经济性的核心构件。对RPV的完整性评估而言辐照脆化是必须面对的问题。我国已开发了第三代设计寿命为60 a的核电站。当达到寿期末时,辐照脆化的行为是未知的,这给国产RPV的辐照脆化预测带来了困难。为研究高注量下的辐照脆化行为,对A508-3钢的材料力学性能试样进行辐照考验,辐照温度为(288±8) ℃,中子注量水平达到反应堆压力容器60 a寿期末的辐照水平1×10²⁰ cm^-2;开展拉伸、冲击和断裂韧性试验,分析辐照脆化行为,在EONY模型基础上,提出针对国产RPV钢的改进的辐照脆化模型。模型的有效性被试验数据证实,其可准确预测国内A508-3材料的辐照脆化趋势。     

低铜压力容器钢辐照脆化效应实验研究

研究了快中子注量率、注量和辐照温度等辐照参数对低铜压力容器钢的辐照脆化程度的影响，从而将实验堆辐照试验数据与动力堆监督试验数据关联。采用了仪表化冲击试验设备和双曲正切函数回归计算的数据处理方法，因而确保了实验结果的准确性。应用半经验公式将仪表化冲击试验数据转化为动态断裂韧性。为压力容器使用寿命评估和新建核电站压力容器设计提供了材料辐照脆化数据。     

反应堆压力容器材料辐照脆化机理研究进展

反应堆压力容器（RPV）材料辐照脆化机理的研究是提高材料辐照脆化抗力、解释辐照脆化效应、建立辐照脆化预测模型的理论基础。开展RPV材料辐照脆化机理的研究不仅有助于认识辐照脆化现象的本质,建立科学的辐照脆化预测模型,改进RPV材料的成分设计和制造工艺,也有助于提高材料的辐照脆化抗力,对于改进RPV材料的性能具有重要意义。本文从RPV材料的发展和微观结构观测手段的进步两方面论述了RPV材料辐照脆化机理研究的两个发展阶段及其主要成果,并对今后的研究手段及研究方向进行了讨论。     

VVER-1000型机组反应堆压力容器辐照脆化评价

反应堆压力容器(RPV)辐照监督及辐照脆化评价是保证核反应堆寿期内安全运行的重要手段。介绍了田湾核电站1、2号VVER-1000型机组辐照监督组件的设置和试验内容,并对1号机组1Л、2Л辐照监督组件的试验结果、辐照脆化预测模型和超前因子进行了分析讨论。结果表明,田湾核电站1号机组RPV母材和焊缝的辐照脆化效应均在原设计标准的范围内,RPV实际辐照脆化趋势与预测模型具有较好的一致性。建议下一套辐照监督组件的抽取时机为运行后第20 a。     

低铜合金反应堆压力容器钢辐照脆化预测评估模型

反应堆压力容器（RPV）材料辐照脆化预测评估对保证核反应堆安全运行、预防重大灾难性事故的发生具有重要意义。通过深入了解RPV材料辐照损伤机理和分析国外较为成熟的RPV辐照脆化预测模型,揭示了国外有关压力容器辐照脆化预测模型对低铜RPV辐照脆化预测的不足及其原因。在此基础上,发展和建立了适用于低铜RPV辐照脆化趋势的预测模型CIAE-2009。利用辐照性能数据对CIAE-2009模型进行了验证。结果表明,CIAE-2009对低铜含量RPV材料辐照脆化趋势预测具有较高的准确性和可靠性。     

某核电厂RPV辐照脆化性能预测与延续运行评估

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。     

利用神经网络方法研究合金元素对反应堆压力容器钢辐照脆化的影响

反应堆压力容器（RPV）作为压水堆中不可更换的关键部件之一,其安全和稳定是决定反应堆安全经济运行的重要因素。RPV钢的辐照脆化问题是制约RPV在堆内安全服役的关键。RPV钢的辐照脆化与其合金成分关系密切。本文利用神经网络方法研究了RPV钢中关键合金成分（Cu、Mn、Ni、Si、P）与辐照脆化之间的关系。研究结果表明,基于神经网络方法得到合金成分与辐照脆化的关系与传统认知基本一致,辐照脆化对Cu含量最敏感,Cu-Ni对辐照脆化存在协同作用,低Cu合金中Mn-Ni、Ni-Si对脆化存在协同作用。     

The development of radiation embrittlement models for US power reactor pressure vessel steels

A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters-Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature - are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.     

Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors

A comparison between pearlitic 2CrMoV steel (WWER-440) and 9% Cr based ferritic-martensitic steels (EUROFER 97 and LA12TaLC) is presented as regards irradiation induced ductile-brittle transition temperature shifts. For neutron doses of 1.5-2 dpa and irradiation temperatures around 300 °C the transition temperature shifts for WWER-440 steel and EUROFER 97 welds are comparable. In the temperature range 350-500 °C the radiation embrittlement levels of both steels are low. Moreover, post-irradiation annealing is proposed as a promising method to predict results of high temperature irradiation embrittlement from existing lower temperature irradiation embrittlement data.     

Effects of Alloy Elements on Irradiation Embrittlement of Reactor Pressure Vessel Steel: Insights from Neural Network Modelling

As one of the key components that can not be replaced in PWR, the safety and stability of reactor pressure vessel (RPV) steel determine the safety and economy of the reactor. The irradiation embrittlement of RPV steel is the limiting factors for the operation of PWR. The irradiation embrittlement of RPV steel is closely related to its alloy composition. Based on the machine learning method, the relationship between key alloy components (Cu/Mn/Ni/Si/P) and irradiation embrittlement of RPV steel was constructed. The results show that the relationship between the alloy composition and irradiation embrittlement is basically consistent with the traditional cognition. The irradiation embrittlement is sensitive to Cu content, and Cu-Ni has synergistic effect on irradiation embrittlement. In low Cu alloys, Mn-Ni and Ni-Si have synergistic effects on embrittlement.     

核压力容器钢冲击断口剪切面积百分比的估算方法

在核压力容器钢的中子辐照脆化评价中,断口剪切面积百分比是一个重要的参数。但此参数不易直接测量,对于辐照后的放射性试样其测量更加困难。本文采用Charpy-V示波冲击试验,并根据计算机采集得到的完整记录冲击过程的载荷－位移曲线,即可确定相应的载荷特征值,同时估算出断口剪切面积百分比。该估算方法用于计算核压力容器钢因中子 辐照引起的脆性转变湿度的变化值,其计算结果较为准确且计算方法也简便,现已成功地应用于大亚湾核电站压力容器的辐照监督试验。     

Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

Irradiation damage caused by neutron irradiation below 425-450 °C of 9-12% Cr ferritic/martensitic steels produces microstructural defects that cause an increase in yield stress. This irradiation hardening causes embrittlement observed in a Charpy impact test as an increase in the ductile-brittle transition temperature. Little or no change in strength is observed in steels irradiated above 425-450 °C. Therefore, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study, significant embrittlement was observed in F82H steel irradiated at 500 °C to 5 and 20 dpa without any change in strength. Earlier studies on several conventional steels also showed embrittlement effects above the irradiation-hardening temperature regime. Indications are that this embrittlement is caused by irradiation-accelerated or irradiation-induced precipitation. Observations of embrittlement in the absence of irradiation hardening that were previously reported in the literature have been examined and analyzed with computational thermodynamics calculations to illuminate and understand the effect.     

Zr－2合金在化学去污液中的辐照吸氢试验

本文叙述了Ｚｒ－２合金在４种去污液中及冷停堆γ辐照场中的吸氢试验。Ｚｒ－２合金经氢分析及金相观察表明：在试验条件下，产生的辐解氢对Ｚｒ－２包壳材料的吸氢无明显影响，更不会导致氢蚀及氢脆。