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。     

提高核压力容器中子辐射脆化断裂韧性监测可靠性的新方法

为了从经受中子辐射过的小样品中获得更多的信息,提高监测试验的可靠性,对断裂韧性试验用过的试样加以二次开发利用具有重要意义。本研究对断裂韧性辐射脆化监测试验用过的 Charpy尺寸试样进行重新设计和利用,从1个试样只能得到1—3个断裂韧性试验数据提高到9个数据,从而显著改善和提高核压力容器中子辐射脆化断裂韧性监测试验和评定的可靠性。同时,对影响预制疲劳裂纹深侧槽 Charpy 尺寸复合试样合理设计的诸因素进行了探讨。     

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     

不同方法评定核压力容器用A508CL3钢动态断裂韧性的比较

采用示被冲击试验及预制疲劳裂纹Ｃｈａｒｐｙ试样，通过几种不同的单试样试验方法对核压力容器用Ａ５０８ＣＬ３钢的动态断裂韧性进行了评定，比较了各种方法的优缺点。研究载荷能量法过高地评定了材料的动态裂韧性值；而柔度变化度法的评定值则过低，断口延伸带宽度测试法的评定结果具有较大误差，且试验过程复杂，成本较高；能量修正法是一种方便而精确的评定核压力容器钢动态断韧性的单试样方法。     

核压力容器钢辐照后动态断裂韧性测试及研究

应用预裂纹示波冲击法研究压力容器钢辐辐照后动态断裂韧性Ｋｉｄ测试方法,结果表明：预裂纹试样的裂纹扩展始于其最大载荷之前,辐照对动态断裂韧性△Ｔ１００的影响小于冲击韧性转变温度△Ｔ４ＩＪ的影响。     

用Chaypy试样研究核压力容器钢的韧性转变温度曲线

用一个标准Ｃｈａｒｐｙ试样及试验过的该试样的两个半截试样所获得的８个“再造Ｃｈａｒｐｙ复合试样”评定核压力容器用Ａ５０８ＣＬ３钢冲击韧性或断裂韧性的温度转变曲线。研究结果表明，用这种方法可得到一条较为可靠而完整的核压力容器钢的冲击韧性或断裂温度的转变曲线。     

用Charpy试样研究核压力容器钢的韧性转变温度曲线

用一个标准Ｃｈａｒｐｙ试样及试验过的该试样的两个半截试样所获得的８个“再造Ｃｈａｒｐｙ复合试样”评定核压力容器用Ａ５０８ＣＬ３钢冲击韧性或断裂韧性的温度转变曲线。研究结果表明，用这种方法可得到一条较为可靠而完整的核压力容器钢的冲击韧性或断裂韧性随温度的转变曲线。这相当于在核监测计划中，只需取出一个Ｃｈａｒｐｙ试样即可得到完整的材料性能曲线，这对于提高核压力容器中子辐照脆化监测的可靠性很有价值，特别是对于运行时间较长（堆芯内监测试样数量已经不多）的核容器寿命预测和安全监视尤为重要。     

反应堆压力容器简体60年寿命疲劳裂纹扩展分析

利用ANSYS程序,对反应堆压力容器筒体在正常运行工况下进行疲劳裂纹扩展分析,获得了反应堆压力容器筒体在60年寿期末的疲劳裂纹尺寸,按照RCC-M规范的要求,对压力容器在主管道破裂瞬态和主蒸汽管破裂瞬态下进行了快速断裂评价。研究结果表明,压力容器满足RCC-M规范的要求,不会发生裂纹失稳。     

Beremin模型预测尺寸效应对A508-3钢断裂韧性的影响

断裂韧性是用于表征反应堆压力容器（RPV）钢脆性状态的重要指标。在开展相关研究时,由于辐照空间小等原因,一般采用小尺寸紧凑拉伸（CT）试样。为掌握CT试样尺寸变化对国产RPV钢断裂韧性测试结果的影响,对国产A508-3钢的不同尺寸CT试样进行了测试分析,采用Beremin模型方法研究了尺寸效应对断裂韧性数据的影响,并建立了不同尺寸CT试样的断裂韧性数据归一化模型（TSM）。结果表明,同一温度下实验测得的断裂韧性值随试样尺寸的减小逐渐增大,不同样品通过标准方法得到的归一化数据存在偏差,本文建立的TSM可有效减小换算数据偏差。     

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.     

Heavy-section steel technology program: Intermediate-scale pressure vessel tests

Eight flawed 990-mm-OD, 152-mm-thick vessels of ASTM A508 class 2 forging steel or A533, grade B, class 1 steel plate have been pressurized hydraulically to burst or rupture. The rupture test of one vessel (V-7) was repeated pneumatically to study the effects of the sustained load thus attainable. Test temperatures ranged from 0 to 91°C so that transitional to upper shelf fracture toughness behavior was observed. Pretest predictions of failure pressure were based on tensile and toughness properties determined from Charpy-size to 102-mm-thick specimens representative of the test vessels. Linear elastic fracture mechanics based on strain and plastic instability analyses were generally adequate for determining the failure pressure, which ranged from 2.15 to 3.28 times design pressure.     

Dynamic fracture toughness test and master curve method analysis of IAEA JRQ material

IAEA conducted a round-robin fracture test program to test and verify the Master Curve method. One of the materials selected for the round-robin is a A-533B1 plate designated as reference material JRQ. Unnotched Charpy-size specimens were fabricated and distributed to a number of testing laboratories. The three US Owners Groups received specimens for both Charpy impact and three-point bending tests to establish fracture toughness master curves. The B&W Owners Group elected to perform a dynamic fracture toughness test under a high loading rate using the JRQ specimens. The master curve method was successfully applied to numerous fracture toughness data sets of pressure vessel steels. Joyce [Small Specimens Test Technique, ASTM STP 1329, 1997, ASTM] applied this method to high loading rate fracture toughness data for A-515 steel and showed the applicability of this approach to dynamic fracture toughness data. This paper presents the data and the resulting reference temperature shift in the Master Curves from static to dynamic fracture data. Based on earlier PTS analyses performed in 1985, an appropriate T₀ shift value is selected for nuclear power plant applications. This shift in T₀ is compared with the temperature shift between K_Ic and K_Ia curves in ASME Boiler and Pressure Vessel Code.     

Mechanism based evaluation of materials behavior and reference curves

The safety assessment of nuclear pressure vessels and piping requires a quantitative estimation of defect growth by stable and unstable manner during service. This estimation is essential for determining whether the defect detected during inspection should be repaired or whether the size of the defect even after its expected growth is small enough to leave the integrity of the vessel unaffected.The most important stable defect growth mechanism is that of environmentally assisted cyclic crack growth. Recent results indicate that it is markedly affected by sulfur content and/or manganese sulfide morphology and distribution. This implies that an essential improvement in component safety has been gained by currently applied steelmaking practices, which result in extra low sulfur content, generally below 0.01 wt%, and in round shape and small size of inclusions, through, e.g., calcium treatment, hence considerably reducing the effect of environment on crack growth rate. This further implies that the ASME Section XI reference curves for environmentally accelerated cyclic crack growth are conservative for steels produced by current steelmaking practices.The ASME Section XI applies predominantly linear elastic fracture mechanics to assess the effects of cracks on the integrity of nuclear power plant components. Unstable linear elastic fracture often propagates by cleavage mechanism. The cleavage fracture process has recently been shown to be of statistical nature in both ferritic and bainitic steels. The carbide size distribution plays a dominant role in controlling the fracture toughness of these steels. A cleavage fracture model has been developed, based on carbide induced cleavage fracture in ferritic and bainitic steels, which can be used to estimate the expected value and probability limits of fracture toughness. This method has been utilized to evaluate the conservatism of the ASME reference fracture toughness curve. For this purpose a microstructural analysis was carried out for the HSST-02 plate material, with which a large amount of K_Ic data has previously been generated for reference curve purpose. The result of the statistical evaluation indicates that based on the 95% survival probability limit some parts of the ASME reference fracture toughness curve are unconservative.     

Moment modified compact tension specimen for measuring crack arrest toughness

Crack arrest toughness in reactor vessel steels in the transition and Charpy upper shelf energy temperature range are of particular interest to the nuclear industry to aid with the analysis of the phenomenon known as pressurized thermal shock (PTS). A test specimen and analysis technique have been developed to measure crack arrest toughness at temperatures from the transition region up to and beyond the Charpy upper shelf energy level. The moment modified compact tension (MMCT) specimen combines a thermal gradient with mechanical loadings to initiate a crack in brittle material below NDT and then have arrest take place in hot, ductile material. A finite element model was used to help design the specimen and fixturing geometry as well as calculate the arrest toughness. Tests have been conducted on ASME SA533 Grade B Class 1 steel plate with a variety of loadings confirming the veracity of the technique and developing valuable data. Crack arrest toughness has been measured from 0°F to 110°F (−18°C to 43°C). This work has been part of a research program performed by C-E, Windsor and funded by the Electric Power Research Institute.     

Subject index to volume 52

Recent experiments on pressure vessel steels have shown that the fracture toughness can be significantly increased at low temperatures if the material is prestressed at a higher temperature. A conservative method is formulated to use this warm prestressing effect in the fracture mechanics analysis of nuclear pressure vessels under thermal shock. This method uses the basic premise that a crack will not initiate when the stress intensity factor is dropping with time (or constant), whether or not the temperature is dropping. A considerable amount of supportive experimental information is presented and a thorough justification of the method is given. One example is presented to illustrate the beneficial aspects of warm prestressing during a thermal shock. The results show that, when warm prestressing is used, the minimum initiation crack depth can be greater, in some cases, by a factor of ten than that calculated by standard methods.