Compared Modeling Study of Primary Water Stress Corrosion Cracking at Dissimilar Weld of Alloy 182 of Pressurized Water Nuclear Reactor According to Hydrogen Concentration
Omar F. Aly*, Miguel M. Neto*, Mônica M. A. M. Schvartzman**, and Luciana I. L. Lima***
*IPEN-Energy and Nuclear Research Institute
2242 Av. Lineu Prestes, S ao Paulo 05508-000, Brazil
**CDTN-Nuclear Technology Development Center, Belo Horizonte, Brazil
***Vallourec Research and Development-Corrosion, Belo Horizonte, Brazil
One of the main failure mechanisms of pressurized water reactors (PWR) is primary water stress corrosion cracking (PWSCC), which occurs in alloy 600 (75Ni-15Cr-9Fe) and weld metals such as alloy 182 (70Ni-14Cr-9Fe), and alloy 82 (73Ni-19Cr-2Fe). Corrosion cracking is due, for example, in reactor nozzles welded dissimilarly with alloys 182/82 between ASTM A-508 G3 steel and AISI316L stainless steel. Corrosion cracks can cause problems reducing nuclear installations safety and reliability. Hydrogen dissolved into primary water to prevent radiolysis, also may enhance PWSCC growth. This article begins from a study by Lima et al. (2011) based on experimental data from the CDTN-Brazilian Nuclear Technology Development Center, and related to a slow strain rate test (SSRT). This was prepared and used for testing welds in the laboratory, similar to the dissimilar weld in pressurizer relief nozzles operating at the Brazilian Angra Unit 1 nuclear power plant. It was simulated for tests, primary water at 325°C and 12.5 MPa containing four levels of dissolved hydrogen. Our objective in this article is to clarify, and discuss adequate modeling based on the SSRT experimental results, and to compare them with those from another database and modeling, of the PWSCC growth rate based on levels of dissolved hydrogen.
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