A508-Ⅲ钢小尺寸试样的拉伸尺寸效应及其归一化模型研究

设计了系列A508-Ⅲ钢小尺寸试样,开展了不同热处理温度和不同厚度下小尺寸试样的室温拉伸测试,分析了小尺寸试样尺寸效应的影响机理。结果表明,随着奥氏体化温度由900℃升高到1 000℃,A508-Ⅲ钢的晶粒尺寸由18.12μm增加到33.21μm;不同热处理温度和不同厚度A508-Ⅲ钢小尺寸试样的拉伸力学性能呈现明显的尺寸效应;随着晶粒尺寸的细化,产生了晶粒尺寸效应,而随着小尺寸试样厚度的减小,产生了特征尺寸效应。同时,引入了能综合描述小尺寸试样晶粒尺寸效应和特征尺寸效应的关键参数λ,构建了考虑尺寸效应的力学本构模型,获得了小尺寸试样与标准试样的屈服强度和抗拉强度归一化模型,并基于试验结果验证了归一化模型的准确性和可靠性,以期为力学性能的换算提供借鉴。     

不同尺寸CT试样的约束度研究

<正>断裂韧性是评价反应堆材料服役性能的关键指标,紧凑拉伸(CT)试样是常用的断裂韧性试样。在反应堆材料的断裂韧性研究中,需使用小尺寸CT样品,以解决因标准试样尺寸过大造成的一系列问题(如辐照后样品少、辐照参数梯度大等)。但CT样品尺寸的变化会改变其约束度,从而影响CT样品的测试数据。因此,为获得有效的小CT样品测试数据,有必要开展不同尺寸CT的约束度研究,掌握约束度变化对断裂韧性测试结果的影响规律,从而实现小尺寸CT试样测试结果     

国产A508-3钢辐照前后小样品断裂韧性测试分析

国产A508-3钢是反应堆压力容器(RPV)用钢,属于低合金铁素体钢,这类材料具有明显的韧脆转变行为,并且在经受中子辐照后,产生明显的辐照脆化效应,降低材料韧性,增加脆性断裂的风险。为掌握中子辐照对压力容器钢断裂韧性的影响,本文研究并掌握了国产A508-3钢0.5CT样品断裂韧性测试技术,并对辐照前后断裂韧性数进行比较,分析了中子辐照对A508-3钢断裂韧性的影响。     

国产反应堆压力容器用16MND5钢的蠕变损伤本构模型研究

为了获得反应堆压力容器（RPV）材料在高温下的蠕变行为,保证RPV在严重事故工况下的完整性,本研究对国产RPV用16MND5钢的高温蠕变性能进行了测试,获得了600~900℃下材料的蠕变性能,并基于应变强化的基本蠕变本构模型与基于延性耗竭理论的蠕变损伤模型,建立了适用于16MND5钢的蠕变损伤本构模型,给出了材料的蠕变损伤模型参数。结果表明,本文提出的蠕变损伤本构模型的有限元模拟数据与试验数据符合性较好,验证了此蠕变损伤模型的正确性。该方法可用于严重事故情况下RPV的蠕变损伤分析,为RPV的完整性分析提供支持。      

载荷分离法对国产A508-Ⅲ钢断裂韧性的适用性测试

利用载荷分离规则化方法对国产A508-Ⅲ钢1/2T-FFCT试样的断裂韧性进行了测试,得到了国产A508-Ⅲ钢的J-R阻力曲线及断裂韧性JQ值,并采用ASTM E1820及GB/T21143标准对结果进行了判定;同时对其中一个辐照后参考转变温度(T_0)测试的断裂韧性数据采用规则化法进行了处理,研究了载荷分离法对国产A508-Ⅲ钢的断裂韧性测试的适用性。     

基于W型试样确定A508-Ⅲ钢参考温度T0的方法研究

在我国早期开发的W型试样的基础上,参考ASTM E1921标准,开发了基于W型试样的断裂韧性测试技术,建立了包括断裂韧性计算、数据有效性判定和参考温度T₀计算等在内的数据分析方法。在-100～-40 ℃下开展了国产A508-Ⅲ钢的W型试样和标准1C(T)试样的测试分析和试验数据的有效性评价。结果表明,基于W型试样可得到满足ASTM E1921标准的有效断裂韧性数据,W型试样数据点均在标准1C(T)试样master curve(主曲线)的置信区间内,基于W型试样确定的参考温度T₀与标准1C(T)试样的非常接近,W型试样可成为RPV辐照监督备选试样。     

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。     

用小尺寸R－CT试样研究核压力容器钢断裂韧性及与Charpy试样的比较

提出了一种用双边带深侧槽的小尺寸圆形紧凑拉伸试样评定核压力容器（ＲＰＶ）钢断裂韧性的单试作试验方法，给出了用该方法测定的两个厂家生产的核压力容器用Ａ５０８ＣＬ３钢的断裂韧性参数，还与Ｃｈａｒｐｙ试样的试验结果及大尺寸标准试样的试验结果进行了比较。研究结果表明：用双边带深侧槽的小尺寸Ｒ－ＣＴ试样测得的断裂韧性值比相同恻槽深度预制疲劳裂纹Ｃｈａｒｐｙ试样的测试值更接近有效断裂韧性值，所以，用于核压力容器断裂韧性的监测是可行的。     

用小尺寸R—CT试样研究核压力容器钢断裂韧性及与Charpy试样的比较

提出了一种用双边带深侧槽的小尺寸圆形紧凑拉伸试样评定核压力容器钢断裂韧性的单试样试验方法，给出了用该方法测定的两个厂家的核压力容器用Ａ５０８ＣＬ３钢的断裂韧性参数，还与Ｃｈａｒｐｙ试样的试验结果及大尺寸标准试样的试验结果进行了比较，研究结果表明，用双边带深侧槽的小尺寸Ｒ－ＣＴ试样测得的断裂韧性值比相同侧槽深度预制疲劳裂纹Ｃｈａｒｐｙ试样的测试值更接近有效断裂韧性值，所以，用于核压力容器断裂韧性的监     

用不同试样评定核压力容器用A508CL3钢动态断裂韧性的比较性研究

通过示波冲击试验，采用预制疲劳裂纹的半尺寸Ｃｈａｒｐｙ试样及标准Ｃｈａｒｐｙ试样评定了核压力容器用Ａ５０８ＣＬ３钢的动态断裂韧性，研究了试样尺寸对动态断裂韧性的温度转变特性的影响。研究结果表明，半Ｃｈａｒｐｙ尺寸试样在低温下较Ｃｈａｒｐｙ试样过高地估计了Ａ５０８ＣＬ３钢的动态断裂韧性，而在上平台温度以上稍低估了Ａ５０８ＣＬ３钢的动态断裂韧性，所评定出的动态断裂韧性的韧／脆转变温度也明显低于标准尺寸     

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。     

基于小冲杆实验方法的国产A508-3钢辐照前后的断裂韧度测定

建立了辐照前国产A508-3钢断裂韧度和小冲杆实验冲压断裂能之间的线性关系,利用该关系和辐照后小冲杆实验冲压断裂能计算得到了辐照后材料的断裂韧度。用Master曲线方法分别处理中子辐照前、后材料的断裂韧度实验数据,得到参考温度t₀。     

Fracture assessment of shallow-flaw cruciform beams tested under uniaxial and biaxial loading conditions

A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow surface flaws. Matrices of cruciform beam tests were developed to investigate and quantify the effects of temperature, biaxial loading, and specimen size on fracture initiation toughness of two-dimensional (constant depth), shallow, surface flaws. The cruciform beam specimens were developed at Oak Ridge National Laboratory (ORNL) to introduce a far-field, out-of-plane biaxial stress component in the test section that approximates the nonlinear stresses resulting from pressurized-thermal-shock or pressure–temperature loading of an RPV. Tests were conducted under biaxial load ratios ranging from uniaxial to equibiaxial. These tests demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for an RPV material. The cruciform fracture toughness data were used to evaluate fracture methodologies for predicting the observed effects of biaxial loading on shallow-flaw fracture toughness. Initial emphasis was placed on assessment of stress-based methodologies, namely, the J–Q formulation, the Dodds–Anderson toughness scaling model, and the Weibull approach. Applications of these methodologies based on the hydrostatic stress fracture criterion indicated an effect of loading-biaxiality on fracture toughness; the conventional maximum principal stress criterion indicated no effect. A three-parameter Weibull model based on the hydrostatic stress criterion is shown to correlate with the experimentally observed biaxial effect on cleavage fracture toughness by providing a scaling mechanism between uniaxial and biaxial loading states.     

Prediction of lower bound fracture toughness for lifetime evaluation of reactor pressure vessels

The potential damage of embrittlement in service is a very important problem of MnMoNi steels used for the nuclear reactor pressure vessel. A decrease of critical flaw size may occur when embrittlement proceeds. The remaining lifetime of the reactors should be assessed taking into account the embrittlement of the steel paying special attention to the degradation of dynamic fracture toughness. The present study introduces the basic concept of the remaining lifetime assessment. Examined was a small specimen fracture toughness test for measuring the dynamic fracture toughness of nuclear reactor pressure vessel (RPV) steels. The result was applied in the measurement of the dynamic fracture toughness of 12 heats of RPV steels. The test results were analyzed to find more practical applications and a method is presented to predict the lower bound dynamic fracture toughness using the Charpy impact test and tensile test results.     

Some issues by using the master curve concept

The master curve concept allows to quantify the variation of fracture toughness with the temperature throughout the ductile-to-brittle transition region. Limit curves of fracture toughness for defined failure probabilities and reference temperatures can be determined using this method. Thus, fracture mechanical values can be supplied for an integrity assessment of structural components. This paper presents the application of the master curve concept to the reference temperature determination over the thickness of the reactor pressure vessel (RPV) steel plate. It was shown that the master curve concept is applicable to the fracture mechanical characterisation of material with different microstructures using small test specimens. The influence of the materials homogeneity and the test temperature on the resulting reference temperature was investigated.     

Characterization of tensile strength and fracture toughness of nuclear graphite NBG-18 using subsize specimens

The mechanical properties of NBG-18 nuclear grade graphite were characterized using small specimen test techniques and statistical treatment on the test results. New fracture strength and toughness test techniques were developed to use subsize cylindrical specimens with glued heads and to reuse their broken halves. Three sets of subsize cylindrical specimens of different sizes were tested to obtain tensile fracture strength and fracture toughness. The mean fracture strength decreased as the specimen size increased. The fracture strength data indicate that in the given diameter range the size effect is not significant and much smaller than that predicted by the Weibull moduli estimated for individual specimen groups of the Weibull distribution. Further, no noticeable size effect existed in the fracture toughness data. The mean values of the fracture toughness datasets were in a narrow range of 1.21-1.26 MPa√m.     

Study on mo steel for high temperature gas-cooled reactor material characterization of forged low Si mo steel

The Japan Atomic Energy Research Institute (JAERI) has carried out a series of research and development work related to the high temperature gas-cooled reactor (HTGR) and, accordingly the high temperature engineering test reactor (HTTR) will be constructed in the near future. As the reactor pressure vessel (RPV) material, Mo steel will be used. Material characterization tests have been carried out to evaluate the applicability of the Mo steel for the RPV and to prepare for the licensing. The present paper summarizes the fracture toughness behavior including K_Id and K_Ia, irradiation embrittlement susceptibility and degradation of steel due to the long term aging at high temperature of the forged low Mo steel. These tests reveal good fracture toughness which well meets the requirements of the ASME Code, low neutron irradiation embrittlement susceptibility, little embrittlement by long term aging and so on. The present test results demonstrate good applicability of forged low Mo steel to the RPV of HTGR.     

基于主曲线方法研究加载速率对RPV材料断裂韧性的影响

利用3组不同材料预制的裂纹夏比试样（PCCv)研究了不同温度和不同加载速率对反应堆压力容器材料断裂韧性的影响，对采集到的实验数据用ASTM E1921—97标淮，计算出材料度的断裂韧性值和参考温度(T0)。从稳态到瞬态加载条件下的参考温度幅值用主曲线方法确定。研究结果表明，T0依赖于加载速率，并随加载速率的增大而增加，同时当稳态加载时的T0较小时，瞬态加载时的T0增值（△T0)较大。