Materials Committee

Mission

The mission of the PWROG Materials Committee (MSC) is to support the Nuclear Promise by providing a forum for the identification, resolution and prevention of materials issues to improve the safe and efficient operation of PWR plants consistent with the objectives of NEI 03-08.

Vision

The vision of the MSC is to be a valued PWR materials leadership group that collaborates internally and with other industry issues programs to support ongoing plant maintenance, repair and mitigation needs, and the long-term operation of the PWR fleet.

Core Teams/Focus Areas

Reactor Internals
The focus of the work is on the plant-specific implementation of the industry requirements to support aging management of Reactor Internals (RI) in PWRs. Research and development-related activities of the RI industry planning activities are typically designated to the MRP as managed and aligned with programs administered by EPRI. Focus areas under the PWROG MSC relative to reactor internals include:

• Responding to identified degradation during RI inspections
• Updating of the reactor internals acceptance criteria methodology to support MRP-227
• Development of industry guidance as needed
• Justification of a 12-Year Re-Inspection Interval for MRP-227 Components
• Development of acceptance criteria to support continued operation if degradation is identified during reactor internals inspections
• Development of methodologies to reduce uncertainty in fluence models for reactor vessel beltline and internals components

Core Barrel Weld Degradation
The focus of this work is on the understanding of the technical issues associated with core barrel cracking. A focus group was established to coordinate industry activities to resolve outstanding issues on potential degradation of the core barrel welds and the impact to plant operations. Focus areas under the MSC relative to core barrel weld degradation include:

• Development of a fault tree for potential causes of cracking
• Development of industry guidance as needed
• Investigation of the effect of the manufacturing process on the potential for flaws adjacent to the core barrel outlet nozzles
• Utilization of neutron noise data to monitor for active core barrel cracking and separation
• Development of PWR core barrel Lower Girth Weld (LGW) failure operability feasibility assessment
• Investigation and exploration of core barrel repair options
• Development of a standard set of welded inspection specimen designs for UT technique development and demonstration purposes

Reactor Vessel Integrity
The focus of the work is on the plant-specific implementation of the industry requirements to support aging management of Reactor Vessel (RV) integrity in PWRs. Research and development-related activities of the RV industry planning activities are typically designated to the MRP as managed and aligned with programs administered by EPRI. Focus areas under the MSC relative to RV integrity include:

• Supporting the qualification/refinement of fluence determination in non-traditional RV beltline locations
• Development of an acceptable method for any licensee to use irradiated fracture toughness data to improve or demonstrate margin in P-T curves
• Development of recommendations for RV surveillance performance monitoring considering 100 years of operation
• Development of the methodology for extending RV in-service inspection interval from 10 to 20 years
• Supporting the review of the proposed changes to 10 CFR 50 Appendix H or the GALL SLR for long-term operation regarding RV surveillance capsule testing requirements and the NRC proposed changes to Regulatory Guide 1.99 Revision 2

Pressure Boundary:
The focus of this work is addressing pressure boundary material issues (e.g., SCC of austenitic stainless steels, degradation of bottom mounted nozzles, and branch connection nozzles). Focus areas under the MSC relative to pressure boundary include:

• Monitoring and proactively addressing the EdF SCC Cracking OE
• Updating industry boric acid guidance for the fleet
• Undertaking efforts to better understand the long-term effects of thermal aging on the pressure boundary components
• Determining the appropriate inspections / repair for Alloy-600 branch connections (i.e. ASME Code Case N-722 or N-770, N-853)