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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

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
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

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.
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    [3] XIAO Zejun, LI Xiang, HUANG Yanping, et al. Nuclear Power Engineering, 2013, 34(1): 1. (in Chinese)(肖泽军, 李翔, 黄彦平, 等. 核动力工程, 2013, 34(1): 1.)
    [4] CHENG Xu, LIU Xiaojing. Atomic Energy Science and Tech-nology, 2008, 42(2): 167. (in Chinese)(程旭, 刘晓晶.原子能科学技术, 2008, 42(2):167.)
    [5] TIM A, SUE I. Energy Policy, 2008, 36: 4323.
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    [7] THOMAS S, LAURENCE K H L, YOSHIAKI O. Progress in Nuclear Energy, 2014, 77: 282.
    [8] LU Daogang, PENG Changhong. Atomic Energy Science and Technology, 2009, 43: 743. (in Chinese)(陆道纲, 彭常宏. 原子能科学技术, 2009, 43: 743.)
    [9] AZEVEDO C R F. Engineering Failure Analysis, 2011, 18:1943.
    [10] GUZONAS D, NOVOTNY R. Progress in Nuclear Energy,2014, 77: 361.
    [11] ARTHUR T M, AYLIN Y, MARCELO J, et al. Journal of Nuclear Materials, 2007, 371: 51.
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    [16] MARIANA R, PETR H, STEPAN S, et al. Nuclear Engi-neering and Technology, 2007, 40(2): 127.
    [17] SCHULENBERG T, STAR? INGER J. Nuclear Engineering and Design, 2011, 241: 3505.
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出版历程
  • 收稿日期:  2016-06-05
  • 修回日期:  2016-09-22
  • 刊出日期:  2017-06-20

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
    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。

  • 中图分类号: TL341

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

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
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