核电SA508-3钢厚壁圆筒纵向焊接残余应力分析

针对由某公司生产首次应用到核电设备上的SA508-3钢,为了获得焊接残余应力分布及规律,采用ANSYS有限元软件对60 mm厚圆筒纵焊的焊接接头进行温度场及残余应力数值模拟,并将模拟结果与相同工艺条件下焊接试验结果进行比较验证.结果表明,模拟结果与试验结果基本吻合;焊接时热源周围极窄区域温度高,梯度大,远离热源温度峰值急剧下降;圆筒外表面残余应力大于内表面残余应力;焊缝及近焊缝区的残余拉应力值较大,远离焊缝中心残余拉应力值逐渐减小;圆筒两端和中部的残余应力在方向上或数值大小上不同;这对控制圆筒残余应力提供了理论依据.     

Fracture toughness and the master curve for modified 9Cr−1Mo steel

Modified 9Cr−1Mo steel is a primary candidate material for the reactor pressure vessel of a Very High Temperature Gas-Cooled Reactor (VHTR) in the Korean Nuclear Hydrogen Development and Demonstration (NHDD) program. In this study, the T₀ reference temperature, J-R fracture resistance and Charpy impact properties were evaluated for commercial Grade 91 steel as part of the preliminary testing for a selection of the RPV material for the VHTR. The fracture toughness of the modified 9Cr−1Mo steel was compared with that of SA508−Gr.3. The objective of this study was to obtain the pre-irradiation fracture toughness properties of the modified 9Cr-1Mo steel as reference data for an investigation of radiation effects. Charpy impact properties of the modified 9Cr-1Mo steel were similar to those of SA508−Gr.3. T₀ reference temperatures were measured as −67.7 and −72.4°C from the tests with standard PCVN (pre-cracked Charpy V-notch) and half-sized PCVN specimens respectively, which were similar to the results for SA508−Gr.3. The K_Jc values of the modified 9Cr-1Mo steel with the test temperatures are successfully expressed by the Master Curve. The J-R fracture resistance of the modified 9Cr−1Mo steel at room temperature was nearly identical to that of SA508−Gr.3; in contrast, it was slightly higher at an elevated temperature.     

核电SA508-3钢大型筒体环焊残余应力分析

为了准确地预测核电SA508-3钢的大型筒体环焊温度和残余应力变化规律,基于ANSYS有限元软件,引入子结构法优化焊接模拟过程,并比较模拟结果与试验数据.结果表明,文中采用的计算方法成功地模拟了更加接近真实的筒体环焊过程,且模拟结果与实测值基本吻合;焊缝上所有节点的热循环曲线相似,热影响区的模拟宽度为4 mm左右;筒体外表面焊缝中心线上任意节点的残余应力大小相近,仅在靠近筒体外表面焊缝中心线附近存在一定的轴向压应力,其它区域大部分为拉应力,环向保持较高的拉应力;筒体焊后产生内凹残余变形;结果可为分析大型筒体环焊残余应力提供参考数据.     

淬火冷速对大型核电压力容器用钢显微组织和力学性能的影响

采用光学显微镜和力学性能测试等研究了淬火冷速对大型核电压力容器用SA508-3钢显微组织及力学性能的影响,尤其对落锤冲击性能的影响。结果表明:随着冷速的增加,SA508-3钢的显微组织由宽大的上贝氏体+粒状贝氏体组织向细小下贝氏体+马氏体组织转变。淬火冷速对SA508-3钢的常/高温强度影响不大,而对冲击韧性的影响显著,尤其对零塑性转变温度(NDTT)的影响显著。低冷速下的NDTT只能达到≥-13.3℃,而高冷速下的NDTT大幅度降低,达到≤-48.3℃。     

核电用钢SA508Gr.4N两段式本构模型的建立及管板终锻火次模拟

采用新一代核电材料SA508Gr. 4N钢的真应力-真应变数据,建立了该材料基于物象的温度、应变及应变速率的两段式流变应力本构模型,引入了相关系数R及平均相对误差AARE,验证本构模型的预测能力,发现相关系数和平均相对误差分别为0. 9915和5. 06%。采用该本构模型进行二次开发,基于Fortran语言编写子程序嵌入DEFORM软件,结合SA508Gr. 4N钢随温度变化的热导率及比热容等实测热物性参数,对核电关键零件管板大锻件的关键终锻火次,包括平锤头展宽、压平凸台及旋转压实3个子工艺进行了系统模拟,分析了在不同工艺参数下,热锻成形过程中的锻件温度场、应力场、等效应变场及最终成形性能,最后得到了合理的锻造工艺方案。     

汽车用先进高强钢韧性断裂模型的研究与应用进展

韧性断裂预测对汽车轻量化产品设计与成形工艺优化有着重要的意义.全面综述了强耦合型与弱(非)耦合型韧性断裂模型的发展与研究现状;重点围绕考虑加载路径与应力状态的断裂失效与成形极限曲线预测、各向异性耦合的失效模型拓展、应变速率与温度效应对材料断裂的影响等3个方面的研究现状及应用效果进行了分析介绍,其中,MMC模型、Lou-...     

Corrosion fatigue crack growth behaviour of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The corrosion fatigue crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was systematically characterized under simulated boiling water reactor normal water and hydrogen water chemistry conditions by low-frequency fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C.In this paper, the observed synergistic effects of environmental, material and loading parameters on the environmental acceleration of fatigue crack growth in low-alloy RPV steels are discussed in the context of the Ford-Andresen model. Additionally, the adequacy and conservatism of the current “ASME XI reference fatigue crack growth curves” of the ASME Boiler & Pressure Vessel Code are critically reviewed and assessed on the basis of the gathered experimental data base and this model. Based on the observed cracking behaviour and the Ford-Andresen model, a simple time-domain superposition model is suggested, which could reduce most of the undue conservatism and eliminate uncertainties of the existing codes and therefore serve as a basis for the development of improved reference fatigue crack growth curves.