Corrosion Science and Technology

The Corrosion Science Society of Korea (한국부식방식학회)
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Review of Factors Affecting IASCC Initiation of Stainless Steel in PWRs

원자로 내부구조물 균열개시 민감도에 미치는 영향인자 고찰

  • Hwang, Seong Sik (Korea Atomic Energy research Institute, Material Safety Technology Development Division) ;
  • Choi, Min Jae (Korea Atomic Energy research Institute, Material Safety Technology Development Division) ;
  • Kim, Sung Woo (Korea Atomic Energy research Institute, Material Safety Technology Development Division) ;
  • Kim, Dong Jin (Korea Atomic Energy research Institute, Material Safety Technology Development Division)
  • 황성식 (한국원자력연구원, 재료안전기술개발부) ;
  • 최민재 (한국원자력연구원, 재료안전기술개발부) ;
  • 김성우 (한국원자력연구원, 재료안전기술개발부) ;
  • 김동진 (한국원자력연구원, 재료안전기술개발부)
  • Received : 2021.08.02
  • Accepted : 2021.08.10
  • Published : 2021.08.31
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Abstract

To safely operate domestic nuclear power plants approaching the end of their design life, the material degradation management strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.

Keywords

Acknowledgement

이 논문은 2021년도 산업통상자원부의 재원으로 한국에너지기술평가원의 지원(20191510301140, 해체원전 원자로 내부구조물 베플포머볼트 조사유기 응력부식 균열열화 특성 분석 기술개발)을 받아 수행된 연구입니다.

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