超临界水冷堆燃料包壳材料的辐照损伤研究进展

郑中成; 郭立平; 唐睿

doi:10.11804/NuclPhysRev.34.02.211

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超临界水冷堆燃料包壳材料的辐照损伤研究进展

郑中成, 郭立平, 唐睿

文章导航 > 原子核物理评论 > 2017 > 34(2): 211-218

郑中成, 郭立平, 唐睿. 超临界水冷堆燃料包壳材料的辐照损伤研究进展[J]. 原子核物理评论, 2017, 34(2): 211-218. doi: 10.11804/NuclPhysRev.34.02.211

引用本文:

郑中成, 郭立平, 唐睿. 超临界水冷堆燃料包壳材料的辐照损伤研究进展[J]. 原子核物理评论, 2017, 34(2): 211-218. doi: 10.11804/NuclPhysRev.34.02.211

ZHENG Zhongcheng, GUO Liping, TANG Rui. Research Development of Irradiation Damage on Fuel Cladding Materials for SCWR[J]. Nuclear Physics Review, 2017, 34(2): 211-218. doi: 10.11804/NuclPhysRev.34.02.211

Citation:

ZHENG Zhongcheng, GUO Liping, TANG Rui. Research Development of Irradiation Damage on Fuel Cladding Materials for SCWR[J]. Nuclear Physics Review, 2017, 34(2): 211-218. doi: 10.11804/NuclPhysRev.34.02.211

超临界水冷堆燃料包壳材料的辐照损伤研究进展

doi: 10.11804/NuclPhysRev.34.02.211

1.
武汉大学物理科学与技术学院, 武汉 430072;
2.
中国核动力研究设计院, 成都 610041

基金项目: 国家国际科技合作专项（2015DFR60370）；国家自然科学基金资助项目（11275140，U1532134）

详细信息

作者简介:
郑中成(1988-),男,山东泰安人,在读博士研究生,从事粒子物理与原子核物理研究;E-mail:zhengzc@whu.edu.cn。

中图分类号: TL341

Research Development of Irradiation Damage on Fuel Cladding Materials for SCWR

1.
School of Physics and Technology, Wuhan University, Wuhan 430072, China;
2.
Science and Technology on Reactor Fuel and Materials Laboratory, Nuclear Power Institute of China, Chengdu 610041, China

Funds: International Science & Technology Program of China(2015DFR60370); National Natural Science Foundation of China(11275140, U1532134)

摘要: 超临界水冷堆（SCWR）是第四代核电站的主力堆型之一，高温、高压、超临界水环境下的辐照损伤问题是其燃料包壳材料面临的最大挑战。SCWR燃料包壳候选材料主要包括锆合金、奥氏体不锈钢、铁素体/马氏体不锈钢、镍基合金、ODS合金五大类，奥氏体不锈钢是最有希望的候选材料。介绍了近年来在这个领域国际上的主要研究进展。作者所在团队也对多种SCWR的候选材料进行了辐照损伤研究，包括：镍基合金C-276和718、铁素体/马氏体钢P92、奥氏体不锈钢AL-6XN和HR3C。对AL-6XN的氢离子辐照实验发现，辐照产生的缺陷主要是间隙型位错环，伯格斯矢量为1/3<111>，在较高剂量（5～7 dpa）辐照下，出现空洞肿胀。在氢滞留的影响下，位错环有着独特的演化规律，总结提出了位错环的四阶段演化过程。

The Supercritical Water-cooled Reactor (SCWR) is one of the prior Generation IV advanced reactors. Irradiation damage is one of the key issues of fuel cladding materials which will suffer serious environment, such as high temperature, high pressure, high irradiation and supercritical water. The candidate materials contain zirconium alloys, austenitic stainless steels, ferritic/martensitic stainless steels, Ni-base alloys and ODS alloys. Austenitic stainless steels are the most promising materials. This paper summarized the international researches on irradiation effects in fuel cladding materials for SCWR. The group of authors also has done many researches in this field, including nickel-base alloy C-276 and 718, ferritic/martensitic steel P92 and austenitic stainless steel AL-6XN and HR3C. In AL-6XN austenitic stainless steels irradiated by hydrogen ions, dislocation loops were the dominant irradiation defects. At higher irradiation dose (5～7 dpa), the voids were found. All the dislocation loops were confirmed to be 1/3<111> interstitial type dislocation loops, and four evolution stages of dislocation loops with hydrogen retention were suggested.
- 超临界水冷堆 /
- 燃料包壳材料 /
- 辐照损伤 /
- 中子辐照
Abstract: The Supercritical Water-cooled Reactor (SCWR) is one of the prior Generation IV advanced reactors. Irradiation damage is one of the key issues of fuel cladding materials which will suffer serious environment, such as high temperature, high pressure, high irradiation and supercritical water. The candidate materials contain zirconium alloys, austenitic stainless steels, ferritic/martensitic stainless steels, Ni-base alloys and ODS alloys. Austenitic stainless steels are the most promising materials. This paper summarized the international researches on irradiation effects in fuel cladding materials for SCWR. The group of authors also has done many researches in this field, including nickel-base alloy C-276 and 718, ferritic/martensitic steel P92 and austenitic stainless steel AL-6XN and HR3C. In AL-6XN austenitic stainless steels irradiated by hydrogen ions, dislocation loops were the dominant irradiation defects. At higher irradiation dose (5～7 dpa), the voids were found. All the dislocation loops were confirmed to be 1/3<111> interstitial type dislocation loops, and four evolution stages of dislocation loops with hydrogen retention were suggested.
- supercritical water-cooled reactor /
- fuel cladding material /
- irradiation damage /
- neutron irradiation

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超临界水冷堆燃料包壳材料的辐照损伤研究进展

doi: 10.11804/NuclPhysRev.34.02.211

1. 武汉大学物理科学与技术学院, 武汉 430072;

2. 中国核动力研究设计院, 成都 610041

基金项目: 国家国际科技合作专项（2015DFR60370）；国家自然科学基金资助项目（11275140，U1532134）

作者简介:
郑中成(1988-),男,山东泰安人,在读博士研究生,从事粒子物理与原子核物理研究;E-mail:zhengzc@whu.edu.cn。

收稿日期: 2016-06-05
修回日期: 2016-09-22
刊出日期: 2017-06-20

中图分类号: TL341

关键词:

摘要: 超临界水冷堆（SCWR）是第四代核电站的主力堆型之一，高温、高压、超临界水环境下的辐照损伤问题是其燃料包壳材料面临的最大挑战。SCWR燃料包壳候选材料主要包括锆合金、奥氏体不锈钢、铁素体/马氏体不锈钢、镍基合金、ODS合金五大类，奥氏体不锈钢是最有希望的候选材料。介绍了近年来在这个领域国际上的主要研究进展。作者所在团队也对多种SCWR的候选材料进行了辐照损伤研究，包括：镍基合金C-276和718、铁素体/马氏体钢P92、奥氏体不锈钢AL-6XN和HR3C。对AL-6XN的氢离子辐照实验发现，辐照产生的缺陷主要是间隙型位错环，伯格斯矢量为1/3<111>，在较高剂量（5～7 dpa）辐照下，出现空洞肿胀。在氢滞留的影响下，位错环有着独特的演化规律，总结提出了位错环的四阶段演化过程。

The Supercritical Water-cooled Reactor (SCWR) is one of the prior Generation IV advanced reactors. Irradiation damage is one of the key issues of fuel cladding materials which will suffer serious environment, such as high temperature, high pressure, high irradiation and supercritical water. The candidate materials contain zirconium alloys, austenitic stainless steels, ferritic/martensitic stainless steels, Ni-base alloys and ODS alloys. Austenitic stainless steels are the most promising materials. This paper summarized the international researches on irradiation effects in fuel cladding materials for SCWR. The group of authors also has done many researches in this field, including nickel-base alloy C-276 and 718, ferritic/martensitic steel P92 and austenitic stainless steel AL-6XN and HR3C. In AL-6XN austenitic stainless steels irradiated by hydrogen ions, dislocation loops were the dominant irradiation defects. At higher irradiation dose (5～7 dpa), the voids were found. All the dislocation loops were confirmed to be 1/3<111> interstitial type dislocation loops, and four evolution stages of dislocation loops with hydrogen retention were suggested.

English Abstract

Research Development of Irradiation Damage on Fuel Cladding Materials for SCWR

1. School of Physics and Technology, Wuhan University, Wuhan 430072, China;

2. Science and Technology on Reactor Fuel and Materials Laboratory, Nuclear Power Institute of China, Chengdu 610041, China

Funds: International Science & Technology Program of China(2015DFR60370); National Natural Science Foundation of China(11275140, U1532134)

Received Date: 2016-06-05
Rev Recd Date: 2016-09-22
Publish Date: 2017-06-20

Keywords:

Abstract: The Supercritical Water-cooled Reactor (SCWR) is one of the prior Generation IV advanced reactors. Irradiation damage is one of the key issues of fuel cladding materials which will suffer serious environment, such as high temperature, high pressure, high irradiation and supercritical water. The candidate materials contain zirconium alloys, austenitic stainless steels, ferritic/martensitic stainless steels, Ni-base alloys and ODS alloys. Austenitic stainless steels are the most promising materials. This paper summarized the international researches on irradiation effects in fuel cladding materials for SCWR. The group of authors also has done many researches in this field, including nickel-base alloy C-276 and 718, ferritic/martensitic steel P92 and austenitic stainless steel AL-6XN and HR3C. In AL-6XN austenitic stainless steels irradiated by hydrogen ions, dislocation loops were the dominant irradiation defects. At higher irradiation dose (5～7 dpa), the voids were found. All the dislocation loops were confirmed to be 1/3<111> interstitial type dislocation loops, and four evolution stages of dislocation loops with hydrogen retention were suggested.

郑中成, 郭立平, 唐睿. 超临界水冷堆燃料包壳材料的辐照损伤研究进展[J]. 原子核物理评论, 2017, 34(2): 211-218. doi: 10.11804/NuclPhysRev.34.02.211

引用本文:

郑中成, 郭立平, 唐睿. 超临界水冷堆燃料包壳材料的辐照损伤研究进展[J]. 原子核物理评论, 2017, 34(2): 211-218. doi: 10.11804/NuclPhysRev.34.02.211

Citation: