Eliminating the risk of Alloy 600 Primary Water Stress Corrosion Cracking (PWSCC) in any part of the Reactor Coolant System (RCS) is a major challenge to allow facing Plants aging and even consider a possible life extension.Many of the most sensitive areas of the RCS have already faced PWSCC and remediation was applied, thanks to repairs or main components replacements- Steam Generators Replacement (SGR) against A600 tubing cracks, Reactor Vessel Head Replacement (RVHR) against A600/182/82 CRDM nozzles cracks, Pressurizer replacement (PZRR) against cracks in heaters nozzles, Surge line Dissimilar Weld (DMW) and/or Safety/Relief lines DMWs.Other areas, expected to support longer production time without PWSCC risk, may also become sensitive while considering life extension. Additional inspections are now required by National Authorities in order to secure this assessment.Most of this "new" sensitive areas are located where it is hardly foreseen to replace the affected component (Reactor Vessel, forinstance), in terms of technical feasibility oreconomical consideration.For these areas, Mitigation techniques havebeen developed in order to eliminate the riskof PWSCC. Two basic processes areemployed:1. Lowering the sensitive material tensile stresses (residual and in-service), under a limit where the material, in contact with the primary water, should not become sensitive2. Isolating the sensitive material from the primary water with a coating made from non-sensitive material.The presentation will list and describe the Mitigation processes, already applied, as well as those potentially available, or under development for most of the RCS A600 areas under consideration, which are not presently managed through Repair or Replacement. Examples of worldwide actual applications will be given.The presentation will conclude on the limits of these Mitigation techniques, when-even if economically preferred- they should be replaced by "classical" remediation, when these are technically achievable.
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