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Effect of Uncontrolled Hydrogen Injection on SCC Susceptibility of 304L Steel in High Temperature Water

[+] Author Affiliations
A. L. Medina-Almazán, J. C. Zenteno-Suárez, C. Arganis-Juárez, N. López-García, M. Marín-Almazo

Instituto Nacional de Investigaciones Nucleares, Ocoyoacac, México

G. Galicia-Aguilar

Instituto de IngenierÌa-Universidad Veracruzana, Boca del Río, México

Paper No. PVP2017-65040, pp. V06BT06A032; 7 pages
  • ASME 2017 Pressure Vessels and Piping Conference
  • Volume 6B: Materials and Fabrication
  • Waikoloa, Hawaii, USA, July 16–20, 2017
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 978-0-7918-5800-4
  • Copyright © 2017 by ASME


For operation of existing nuclear power plants (NPPs) beyond their design lifetime (up to 80 years), one of the main issues is the assessment of the performance of its structures, systems and components (SSC) during the period of extended operation. In Boiling water reactors (BWR), the reactor pressure vessel internals (RPVI) components are submitted to both, Neutron irradiation and high oxidizing environment which lead (together, by itself or mixed with a sensitized condition) to stress corrosion cracking (SCC) [1–4]. One of the measures to mitigate SCC in BWR is the decrease of the Electrochemical Corrosion Potential (ECP) of the structural materials by the injection of hydrogen (1-2 ppm H2) and Noble Metal Chemical Addition (NMCA) or On-Line NobleChemTM (OLNC). However, ECP is a thermodynamic measurement that little says about the kinetics of the process of SCC; by the other side, literature shows that hydrogen has an effect on austenitic stainless steels under Light Water Reactors (LWR) environments in corrosion fatigue crack initiation and corrosion fatigue crack growth [4], whilst it has been observed an hydrogen effect by Slow Strain Rate Tests (SSRT) only in stainless steels irradiated to high fluence and in Pressurized Water Reactor (PWR) primary water [4].

The results of Slow Strain Rate Tests (SSRT) performed in a 304L SS at 288°C and 8 MPa are presented in this work. 50% cold worked 304L SS was tested in High Oxygen Water Chemistry (HOWC, no hydrogen injection) and in HWC with Platinum hydrothermal deposits (PtHD); by the other hand, sensitized 304L SS with PtHD was tested under excess hydrogen injection condition (EHI, >8 ppm H2). It was observed that both, 50% cold worked 304L SS tested under HOWC and sensitized 304L SS with Pt deposits tested under excess hydrogen injection, fail in a brittle way, by transgranular stress corrosion cracking (TGSCC), whereas 50% cold worked 304L SS with Pt deposits tested in HWC presents a ductile fracture surface. Even the brittle behavior of the fracture surface of sensitized 304L SS with Pt deposits tested under uncontrolled hydrogen injection, it is not a embrittlement effect on the final elongation f), nevertheless it is seen a hardness increase that could be attributed to H2.

Copyright © 2017 by ASME



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