压水堆一回路环境中304不锈钢的蠕变特性分析

1.西安科技大学 机械工程学院,陕西 西安 710054; 2.西安科技大学 理学院,陕西 西安 710054

304不锈钢; 高应力蠕变; 本构方程; 蠕变行为

Creep characteristics analysis of 304 stainless steel in pressured water reactor primary circuit
XUE He1,CUI Ying-hao1,ZHAO Ling-yan2,TANG Wei1,NI Chen-qiang1

(1.College of Mechanical and Engineering,Xi'an University of Science and Technology,Xi'an 710054,China; 2.College of Sciences,Xi'an University of Science and Technology,Xi'an 710054,China)

304 stainless steel; high stress creep; constitutive equation; creep behavior

DOI: 10.13800/j.cnki.xakjdxxb.2018.0122

备注

裂尖结构蠕变应变是核电结构材料应力腐蚀裂纹扩展的主要驱动力之一,为了了解核电结构材料在核电高温水环境下的蠕变特性,本文利用高压釜模拟核电一回路高温高压水环境,对核电结构材料304不锈钢进行了不同应力下的单轴拉伸蠕变实验,基于时间硬化本构模型得出了其在320 ℃下的蠕变本构方程,并结合ABAQUS有限元分析软件建立了高应力下获取蠕变的数值模拟方法。结果 表明,应力和时间对蠕变变形有着很大的影响,蠕变速率在初期很大,随着蠕变时间的延长,由于合金加工硬化现象的产生,导致蠕变速率逐渐减少并趋于稳定; 温度一定时,蠕变变形和蠕变速率同样随着应力的增大而增大。利用ABAQUS可以有效获取高应力下蠕变规律的数值模拟方法,研究结果为核电结构材料安全性评定提供了一定的参考作用。

Creep strain is one of the main driving forces that cause environmental assisted cracking in nuclear structure material at crack tip.To understand the creep characteristics of nuclear structure materials,Autoclave was used to simulate the high temperature and high pressure water environment in nuclear power circuit,and the uniaxial tensile creep experiment of 304 stainless steel structural materials under different stress was conducted.Based on the creep test data,the creep constitutive model of 304 stainless steel wasbuilt based on time hardening constitutive model,and the creep experiment method was established under high stress using ABAQUS.Results show that stress and time have a great influence on the creep deformation.The initial creep rate is very high,and the degree of work hardening of the 304 increases with the increasing of time,resulting in the creep rate decreases and tends to be stable.The creep deformation and the creep rate increase with the increase of stress under the constant temperature.The creep numerical simulation method of high stress was established using ABAQUS.Results provide a reference for the safety evaluation of nuclear power materials.