当前位置: X-MOL 学术Nucl. Eng. Des. › 论文详情
Our official English website, www.x-mol.net, welcomes your feedback! (Note: you will need to create a separate account there.)
High temperature deformation behavior of Indian PHWR Calandria material SS 304L
Nuclear Engineering and Design ( IF 1.7 ) Pub Date : 2020-11-01 , DOI: 10.1016/j.nucengdes.2020.110801
Keshav Mohta , Suneel K. Gupta , Soupramanien Cathirvolu , Swaminathan Jaganathan , J. Chattopadhyay

Abstract Calandria, a horizontal stepped cylindrical vessel, houses the fuel channels (reactor core) of Indian Pressurized Heavy Water Reactors (IPHWRs). The Calandria and its support assembly is made from austenitic stainless steel SS 304L grade. Post Fukushima nuclear accident (2011), the realistic assessment of time line of structural degradation of Calandria during the progression of a postulated severe accident has become an important requirement. This calls for a detailed finite element analysis of Calandria assembly where material deformation behavior owing to both creep and plasticity up to very high temperatures arising during accident progression has to be modeled. Although stainless steel SS 304L finds wide application in nuclear industry, the material properties up to very high temperatures as required for present case, are not available in the open literature. Hence, in the present work, the tensile and creep-stress rupture properties of SS 304L have been generated for temperatures up to 1100 °C. Engineering stress-strain curves, yield strength, ultimate tensile strength, uniform elongation as well as Ramberg-Osgood fitting parameters are evaluated from the tensile test data. Creep test data is processed to evaluate creep curves, minimum creep rate, Norton-Bailey creep law parameters as well as Larson Miller Parameter based correlation for prediction of stress rupture life. The generated properties would be useful as material inputs to carry out realistic structural integrity assessment of Calandria assembly for loads arising under accident conditions.

中文翻译:

印度PHWR Calandria材料SS 304L的高温变形行为

摘要 Calandria 是一种水平阶梯式圆柱形容器,装有印度加压重水反应堆 (IPHWR) 的燃料通道(反应堆堆芯)。Calandria 及其支撑组件由 SS 304L 级奥氏体不锈钢制成。福岛核事故后(2011 年),在假定的严重事故发展过程中,对排管结构退化时间线的现实评估已成为一项重要要求。这需要对 Calandria 组件进行详细的有限元分析,其中必须对由于蠕变和塑性导致的材料变形行为进行建模,直至在事故进展期间出现非常高的温度。尽管不锈钢 SS 304L 在核工业中得到了广泛的应用,但其材料性能可达到目前所需的非常高的温度,在公开文献中不可用。因此,在目前的工作中,SS 304L 的拉伸和蠕变应力断裂性能已在高达 1100 °C 的温度下产生。工程应力-应变曲线、屈服强度、极限抗拉强度、均匀伸长率以及 Ramberg-Osgood 拟合参数均根据拉伸试验数据进行评估。处理蠕变测试数据以评估蠕变曲线、最小蠕变速率、Norton-Bailey 蠕变定律参数以及基于拉森米勒参数的相关性,以预测应力断裂寿命。生成的属性可用作材料输入,以针对事故条件下产生的载荷对 Calandria 组件进行真实的结构完整性评估。SS 304L 的拉伸和蠕变应力断裂性能已在高达 1100 °C 的温度下产生。工程应力-应变曲线、屈服强度、极限抗拉强度、均匀伸长率以及 Ramberg-Osgood 拟合参数均根据拉伸试验数据进行评估。处理蠕变测试数据以评估蠕变曲线、最小蠕变速率、Norton-Bailey 蠕变定律参数以及基于拉森米勒参数的相关性,以预测应力断裂寿命。生成的属性可用作材料输入,以针对事故条件下产生的载荷对 Calandria 组件进行真实的结构完整性评估。SS 304L 的拉伸和蠕变应力断裂性能已在高达 1100 °C 的温度下产生。工程应力-应变曲线、屈服强度、极限抗拉强度、均匀伸长率以及 Ramberg-Osgood 拟合参数均根据拉伸试验数据进行评估。处理蠕变测试数据以评估蠕变曲线、最小蠕变速率、Norton-Bailey 蠕变定律参数以及基于拉森米勒参数的相关性,以预测应力断裂寿命。生成的属性可用作材料输入,以针对事故条件下产生的载荷对 Calandria 组件进行真实的结构完整性评估。均匀伸长率以及 Ramberg-Osgood 拟合参数由拉伸试验数据评估。处理蠕变测试数据以评估蠕变曲线、最小蠕变速率、Norton-Bailey 蠕变定律参数以及基于拉森米勒参数的相关性,以预测应力断裂寿命。生成的属性可用作材料输入,以针对事故条件下产生的载荷对 Calandria 组件进行真实的结构完整性评估。均匀伸长率以及 Ramberg-Osgood 拟合参数由拉伸试验数据评估。处理蠕变测试数据以评估蠕变曲线、最小蠕变速率、Norton-Bailey 蠕变定律参数以及基于拉森米勒参数的相关性,以预测应力断裂寿命。生成的属性可用作材料输入,以针对事故条件下产生的载荷对 Calandria 组件进行真实的结构完整性评估。
更新日期:2020-11-01
down
wechat
bug