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Grain Boundary Oxidation of Neutron Irradiated Stainless Steels in Simulated PWR Water

  • Takuya FukumuraEmail author
  • Koji Fukuya
  • Katsuhiko Fujii
  • Terumitsu Miura
  • Yuji Kitsunai
Conference paper
Part of the The Minerals, Metals & Materials Series book series (MMMS)

Abstract

To elucidate the mechanisms of irradiation assisted stress corrosion cracking (IASCC), stress corrosion cracking (SCC) tests on 3 dpa, 19 dpa and 73 dpa neutron-irradiated 316 stainless steel were performed and the effects of irradiation on grain boundary (GB) oxidation were investigated. O-ring specimens were prepared from irradiated flux thimble tubes and a constant load SCC test was performed in a simulated pressurized water reactor primary water at 320 °C. After the SCC test, the oxidation condition of GBs was examined by transmission electron microscopy. Evidence of GB oxidation was found in all examined GBs, even at the relatively low dose of 3 dpa. The morphology of GB oxidation was sharp wedge-shaped. The average GB oxidation length at 3 dpa, 19 dpa and 73 dpa were 100 nm, 340 nm and 400 nm, respectively, indicating the promotion of GB oxidation due to irradiation. In the GB oxide, Fe and Ni depletion and Cr enrichment were observed. Also, Ni enrichment on GB was observed in front of the GB oxidation.

Keywords

Irradiation assisted stress corrosion cracking Stainless steel Corrosion PWR Grain boundary 

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Copyright information

© The Minerals, Metals & Materials Society 2019

Authors and Affiliations

  • Takuya Fukumura
    • 1
    Email author
  • Koji Fukuya
    • 1
  • Katsuhiko Fujii
    • 1
  • Terumitsu Miura
    • 1
  • Yuji Kitsunai
    • 2
  1. 1.Institute of Nuclear Safety System, IncFukuiJapan
  2. 2.Nippon Nuclear Fuel Development Co., LtdIbaraki-KenJapan

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