Relative stress corrosion susceptibilities of alloys 690 and 600 in simulated boiling water reactor environments
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INTRODUCTION
THE intergranular stress corrosion cracking (IGSCC) failure of an INCONEL* alloy 600 safe-end forging from the *A trademark of the INCO family of companies.
recirculation inlet nozzle of a boiling water reactor (BWR) in 1978 raised serious questions about the long-term suitability of alloy 600 for use in BWR environments. The results of a failure analysis on the safe-end indicated that the failure was caused by IGSCC occurring under the combined influence of high stresses and environmental conditions associated with a tight crevice between the safe-end and a thermal sleeve welded to the safe-end, l The cracking initiated in a weld heat affected zone and propagated through the weld made with INCONEL Welding Electrode 182 raising questions about its susceptibility and that of welds made with the similar INCONEL Filler Metal 82. This failure indicated a need for further information concerning the IGSCC susceptibility of alloy 600, 1-82, and 1-182 under BWR operating conditions and a possible need for materials to replace them in BWR applications. A proposed replacement material for alloy 600 is the high chromium INCONEL alloy 690 which has thermal and mechanical properties similar to those of alloy 600 and which appears (based on limited test data) to be more resistant than alloy 600 to IGSCC. INCO has also developed high chromium weld alloys, R-127 and R-135, which are compatible with alloy 690 and could be used to replace I-182 and/or 1-82. This paper describes the results of a research program which was conducted to evaluate the IGSCC behavior of alloys 600 and 690 under conditions simulating present operating BWRs. The emphasis of the program was the determination of the relative IGSCC susceptibilities of alloys 600 and 690 and a better definition of the conditions under which the alloys may be susceptible to IGSCC. Results of similar tests on various weld metals are reported elsewhere. 2'3'4 Slow-strain-rate (SSR) tests were used to evaluate the effects of dissolved oxygen content, degree of R.A. PAGE and A. McMINN are Senior Research Engineers, Southwest Research Institute, 6220 Culebra Road, RO. Drawer 28510, San Antonio, TX 78284. Manuscript submitted June 14, 1985. METALLURGICAL TRANSACTIONS A
sensitization, and crevice condition on the IGSCC susceptibility of both alloy 600 and 690 specimens. Crackgrowth tests were conducted concurrently with the SSR tests to obtain data regarding the effects of stress intensity on crack propagation. The primary coolant in a BWR is typically neutral pH, high purity water containing no additives, with radiolysis in the reactor core leading to an oxygen concentration of approximately 200 ppb. Tests were therefore conducted in a high purity water environment to simulate the "normal" primary coolant chemistry. However, water chemistry transients brought about by resin releases from the demineralizer system do occasionally introduce impurities into the primary coolant system. Since demineralizer resins rapidly decompose at normal operating temperatures to yiel
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