Development of maintenance technique for materials of core components in light-water reactor using ion-beam irradiation

轻水堆核心部件材料离子束辐照维护技术开发

基本信息

批准号：
14380230
负责人：
HASEGAWA Akira
金额：
$ 9.28万
依托单位：
TOHOKU UNIVERSITY
依托单位国家：
日本
项目类别：
Grant-in-Aid for Scientific Research (B)
财政年份：
2002
资助国家：
日本
起止时间：
2002 至 2004
项目状态：
已结题

项目摘要

Austenitic stainless steel is used extensively in the core components of light water reactors (LWRs). Irradiation assisted stress corrosion cracking (IASCC) is one of the major concerns of stainless steel in LWR environment. Radiation induced segregation (RIS), that represents chromium depletion and nickel enrichment, caused by neutron irradiation is regarded as one of controlling factors of IASCC. To estimate the effect of chromium depletion at grain boundaries on corrosion behavior, the electrochemical potentiokinetic reactivation (EPR) test is very useful. In order to study the neutron irradiation effect of neutron, energetic particle beam irradiation using accelerators have been performed as simulation experiments. Since the corrosion behavior evaluated by EPR test is affected by chemical properties from the surface to 2-3μm depth of the austenitic stainless steel, the light ion irradiation can introduce enough damage level and damage area to evaluate the corrosion behavior of irra … More diated surface including grain boundaries. The purposes of this study are to establish the testing method using miniature size specimens for ion-irradiation experiments and to clarify the change in corrosion behavior of irradiated stainless steel caused by displacement damage under controlled irradiation conditions such as temperature and fluence.The corrosion behavior of type 304 austenitic stainless steel irradiated with hydrogen ion was investigated using EPR testing combined with TEM observation after irradiation at 300℃, 400℃ and 500℃ up to 0.1, 0.5 and 1 dpa. The results are summarized as follows ;1.It was confirmed that 3 mm disk specimens could be used to estimate corrosion behavior of austenitic stainless steel using EPR test.2.In the specimens irradiated at 300℃ and 400℃, the surface pitting was observed within grains after EPR testing. No etching trenches at and around grain boundaries were observed, nevertheless RIS was occurred in the specimens irradiated at 400℃ to 0.1dpa. So it was suggested that the corrosion resistance at grain boundaries after ion irradiation was still maintained.3.In the specimens irradiated at 500℃, etching trenches at and around grain boundaries were observed, and surface pits were absent. The width of etching trenches and the reactivation ratio increased with irradiation dose. It was suggested that high-temperature irradiation enhanced RIS and also the recovery of damage defects and therefore it suppressed the chemical activation within grains.4.The correlation between the changes in corrosion properties and the development of RIS and microstructures was discussed.Based on these results, it is concluded that hydrogen ion beam irradiation combined with EPR test is valuable to estimate neutron irradiation induced corrosion behavior of austenitic stainless steel. Less

奥氏体不锈钢广泛用于轻水反应堆（LWRS）的核心成分中。辐照辅助应力腐蚀破裂（IASCC）是LWR环境中不锈钢的主要关注点之一。辐射诱导的分离（RIS），代表由中子照射引起的铬消耗和镍富集，被认为是IASCC的控制因素之一。为了估计铬耗尽对晶界对腐蚀行为的影响，电化学预稳定性重新激活（EPR）测试非常有用。为了研究中性的中性辐照效应，已作为仿真实验进行了使用加速器的能量颗粒光束照射。由于通过EPR检测评估的腐蚀行为受到奥氏体不锈钢从表面到2-3μm深度的化学特性的影响，因此轻离子辐照可以引入足够的损伤水平和损伤面积，以评估IRRA的腐蚀行为…更贴发的腐蚀行为，包括晶界。这项研究的目的是使用微型尺寸标本来建立测试方法，用于离子 - 辐射实验，并阐明由受控辐射条件下的位移损伤引起的辐照不锈钢引起的腐蚀行为的变化，例如温度和流体等腐蚀，（例如，使用304型固定型）的腐蚀行为与水力发电的腐蚀行为相结合。 300℃，400℃和500℃最高0.1、0.5和1 dpa。结果总结如下； 1.证实，可以使用EPR测试来估算奥氏体不锈钢的3 mM磁盘样品。2。在EPR测试后在晶粒内观察到在300℃和400°时的样品。未观察到晶界和周围周围的蚀刻沟，但是在400°至0.1DPA的样品中仍发生了RI。因此，有人提出仍保持离子辐照后晶界处的耐腐蚀性。3。在500℃的样品中，观察到晶粒边界和周围的蚀刻沟，没有表面坑。蚀刻沟的宽度和重新激活比随辐照剂量增加。有人提出，高温照射增强了RIS以及损伤缺陷的恢复，因此它抑制了晶粒内的化学激活。4。讨论了腐蚀性的变化与RIS的变化与RIS和微结构的发展之间的相关性。基于这些结果，与EPR iNRADIAD IRRADITIAD IRRADISITION估计相关的易位性易位估计值得估计不锈钢。较少的