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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（LWR）的核心部件。辐照辅助应力腐蚀开裂（IASCC）是不锈钢在轻水堆环境中的主要问题之一。中子辐照引起的辐照诱导偏析（RIS）是IASCC的主要控制因素之一，表现为铬的贫化和镍的富集。电化学动电位再活化（EPR）试验是评价晶界铬贫化对腐蚀行为影响的有效方法。为了研究中子的中子辐照效应，进行了加速器高能粒子束辐照的模拟实验。由于EPR试验评价的腐蚀行为受奥氏体不锈钢表面至2-3μm深度的化学性质的影响，因此，轻离子辐照可以引入足够的损伤程度和损伤面积，以评价奥氏体不锈钢的腐蚀行为。 ...更多信息包括晶界的膨胀表面。本研究的目的是建立一种使用微型试样进行离子-为了阐明在控制辐照温度和能量密度的条件下，位移损伤引起的辐照不锈钢腐蚀行为的变化，采用EPR测试结合透射电镜观察，研究了304奥氏体不锈钢在氢离子辐照后的腐蚀行为。在300℃、400℃和500℃下分别辐照0.1、0.5和1dpa。结果表明：1.采用3 mm圆片试样进行EPR测试，可以较好地评价奥氏体不锈钢的腐蚀行为; 2.在300℃和400℃辐照试样中，EPR测试后在晶内观察到表面点蚀。在400℃ ~ 0.1dpa辐照下，试样晶界处及晶界附近未发现腐蚀沟，但出现了RIS。在500℃下辐照的试样中，在晶界处及晶界附近出现了腐蚀沟槽，表面无凹坑。随着辐照剂量的增加，刻蚀沟槽的宽度和再激活率增加。结果表明，高温辐照促进了RIS和损伤缺陷的修复，从而抑制了晶内化学活化。4.讨论了腐蚀性能变化与RIS和显微组织发展之间的关系。因此，氢离子束辐照结合EPR试验对评价中子辐照腐蚀行为是有价值的。奥氏体不锈钢少