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One of the main degradation mechanisms which cause risks to safety and reliability of pressurized water nuclear reactors is the primary water stress corrosion cracking (PWSCC) in nickel alloys, such as Alloy 600 (75Ni-15Cr-9Fe), and its weld metal Alloy 182 (67 Ni-15Cr-8Fe). It can appear at several reactor nozzles dissimilarly welded with Alloys 182/82 between steel ASTM A-508 G3 and stainless steel AISI316L, among others. The hydrogen which is dissolved to primary water to prevent radiolysis, can also have influence on the stress corrosion cracking behavior. In this article one departs from a study of Lima based in experimental data obtained from CDTN-Brazilian Nuclear Technology Development Center, in slow strain rate test (SSRT). It was prepared and used for tests a weld in laboratory, similar to dissimilar weld in pressurizer relief nozzles, operating at Brazilian NPP Angra 1. It was simulated for tests, primary water at 325 o C and 12.5 MPa containing levels of dissolved hydrogen: 2, 10, 25, and 50 cm 3 STP H 2 /kgH 2 O. The objective of this article is to propose an adequate modeling based on these experimental results, for PWSCC crack growth rate according to the levels of dissolved hydrogen, based on EPRI-MRP-263 NP. Furthermore, it has been estimated the stress intensity factor applied for these tests: according with these, some another models described on EPRI-MRP-115, and an USNRC Technical Report, have been tested. According to this study, CDTN tests are adequate for modeling comparisons within EPRI and USNRC models.

Modeling of Tests of Primary Water Stress Corrosion Cracking of Alloy 182 of Pressurized Water Reactor According to EPRI and USNRC Recommendations