ML18220A742

Summary of the September 12, 2018, Meeting to Discuss Operating Experience and MRP-227, Revision 1 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline.

ML18220A742
Person / Time
Issue date:	10/29/2018
From:	Joseph Holonich NRC/NRR/DLP/PLPB
To:	Dennis Morey NRC/NRR/DLP/PLPB
Holonich J, NRR/DLP, 415-7297
References
Download: ML18220A742 (4)

Text

October 29, 2018 MEMORANDUM TO: Dennis C. Morey, Chief Licensing Processes Branch Division of Licensing Projects Office of Nuclear Reactor Regulation FROM: Joseph J. Holonich, Senior Project Manager /RA/

Licensing Processes Branch Division of Licensing Projects Office of Nuclear Reactor Regulation

SUMMARY

OF THE SEPTEMBER 12, 2018, MEETING TO DISCUSS OPERATING EXPERIENCE AND MRP-227, REVISION 1 MATERIALS RELIABILITY PROGRAM: PRESSURIZED WATER REACTOR INTERNALS INSPECTION AND EVALUATIONS GUIDELINE On September 12, 2018, U. S. Nuclear Regulatory Commission (NRC) staff met with representatives from the Electric Power Research Institute (EPRI), and the Pressurized Water Reactor Owners Group (PWROG). These organizations will be referred to collectively as the industry in this summary.

The purpose of the meeting was to discuss operating experience and Materials Reliability Program (MRP)-227, Revision 1 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline. The objective of the meeting was for EPRI to provide the NRC staff with information on the recent operating experience as it relates to the ongoing review of MRP-227, Revision 1. Information related to the meeting, including presentations, documents discussed in this summary, and the attendees list can be found in the Agencywide Document Access and Management System package ML18206A140.

At the opening of the meeting NRC staff stated that it was working on preparing the safety evaluation (SE) for MRP-227, Revision 1. Continuing, the NRC staff explained that it understood the industry desire to have a clean SE without any conditions. However, the NRC staff explained that it needed to understand if the industry wanted an SE on an accelerated schedule or wanted to wait for the SE until all the information could be evaluated. In response, the industry noted it would like an efficient SE with no conditions.

The NRC staff said that it would prepare the SE on a schedule that would allow it to review all the available information but that, if the staff found MRP-227, Revision 1 acceptable for use, the inclusion or exclusion of conditions in the SE would depend on what the NRC staff needed to document to protect safety.

CONTACT: Joseph J. Holonich, NRR/DLR/DLP 301-415-7297

D. Morey The NRC staff asked about the status of implementation of MRP-227, Rev. 1 by the plants. The industry indicated that most plants are governed by existing license-renewal commitments.

Therefore, the reactor vessel internal (RVI) aging management programs are in accordance with MRP-227-A.

EPRI explained that plants currently using MRP-227-A are implementing it along with guidance that was issued since the submittal of MRP-227-A. Industry noted that if MRP-227, Revision 1 was accepted for use, then its implementation would be analyzed using the methodology in Nuclear Energy Institute guidance 03-08, "Guideline for the Management of Materials Issues."

The NRC staff questioned if plants would have to come to the NRC for any changes coming out of the NEI 03-08 analysis. Industry said that it would conduct an evaluation under Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59, "Changes, tests, and experiments." This analysis was the standard approach for handling changes at plants, and would determine if plants needed to come to the NRC.

Next, the recent EPRI guidance on control rod drive mechanism (CRDM) thermal sleeves was discussed. An NRC staff question dealt with whether impacts to the inside and outside diameters for CRDM thermal sleeves would be addressed in the responses to the Requests for Additional Information (RAIs). Industry responded that the PWROG was working on this topic and that the RAI response would discuss what is being done.

In addition, industry stated that PWROG-16003-P, Evaluation of Potential Thermal Sleeve Flange Wear, was being revised based on the work being done. Industry stated that PWROG-16003-P is intended to be the inspection and evaluation document for thermal sleeves. MRP-227 would then refer to this guidance, similar to the guide card guidance in WCAP-17451-P, "Reactor Internals Guide Tube Wear-Westinghouse Domestic Fleet Operational Projections."

Industry agreed to provide the PWROG-16003-P revision to the NRC staff for information when it is completed. Also, industry suggested that a pre-submittal meeting might be worth holding even if the revision was only being provided for information.

The NRC staff stated that it was conducting an assessment under LIC-504, "Integrated Risk-Informed Decision-Making Process for Emergent Issues." The NRC staff expected that the assessment would be done within approximately two weeks. Therefore, it was suggested that a meeting be held to discuss the NRC staff LIC-504 assessment and results in about a month from this meeting.

The industry indicated it does not plan to incorporate the interim guidance of CRDM thermal sleeves into MRP-227, Revision 1 due to the possibility of changes in this guidance in the near term. Industry does however, plan to issue thermal sleeve inspection and evaluation guidance under the revision to PWROG-16003-P by the first quarter of 2019. It would subsequently incorporate the PWROG guidance for CRDM thermal sleeves into MRP-227, Revision 2, expected to be submitted for NRC review in late 2020.

At the end of the discussions on this topic, industry explained that its overall process was to respond to the RAIs and look for the NRC staff to complete its evaluation of MRP-227, Revision 1. Plants would continue to follow guidance, such as that on the CRDM thermal sleeves, that was developed after the submittal of MRP-227, Revision 1. In addition, industry expected more information on the CRDM thermal sleeves in 2019.

D. Morey This information and the interim guidance would then be incorporated into MRP-227, Revision 2. Also, industry indicated that, although MRP-227, Revision 2 was originally intended for subsequent license renewal (SLR) work for 80 years, the goal is to have it apply to original license life and any life-extension period beyond the original operating license. The industry intent is that MRP-227 is applicable as a single NEI 03-08 inspection and evaluation guidance document for aging management of PWR reactor internals.

The next topic covered in the meeting was core-barrel welds. As part of the interactions on this topic, the NRC staff asked if industry was doing a root-cause analysis of the cracks found in core barrels. Industry responded that it was not doing a root-cause analysis but it was not known if the licensee of the plant was doing one. To respond to the RAI on core-barrel weld operating experience, EPRI would commit to increased coverage in MRP-227, Revision 1.

NRC staff had a presentation related to core-barrel welds. However, given the information provided in the industry presentations, the NRC staff presentation was not as relevant. The changes presented in the industry presentation was an increase in core-barrel weld inspections from the current 25% of the weld length to 100% of the accessible weld length, with 50%

minimum being required. The NRC staff presentation was based on the original inspection number of 25%.

The final industry presentation covered baffle-former bolts (BFBs). In its presentations on baffle bolts, industry said that the appropriate location for the guidance related to BFBs was in the acceptance criteria methodology in WCAP-17096-NP-A, Revision 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements." Industry reported that the plans were to revise WCAP-17096-NP in 2019. The issue would be addressed there, with guidance consistent with the recommendation from the NRC Staff Assessment of the BFB Interim Guidance.

A final topic covered at the meeting was the discussion of Action Item (AI) 8, "Submittal of Information for Staff Review and Approval," from the 2011 staff safety evaluation of MRP-227, Revision 0. The NRC staff explained that the AI was added as a reminder to include certain information related to RVI in the license renewal application. Now, the NRC staff explained, it has determined that the information required by AI 8 is already required by 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants."

The NRC staff also indicated that a requirement in AI 8 to address environmental effects of fatigue for RVIs is not needed because it is inconsistent with most plants current licensing basis and is not consistent with other guidance for license renewal, such as the Standard Review Plan. Because of that, the NRC staff stated that the AI was no longer needed and, if MRP-227, Revision 1 was found acceptable for use, the AI would not be included in the SE.

Docket No. 99902016

Package (ML18006A140); Summary (ML18220A742)

• concurrence via email NRC -001 OFFICE NRR/DLP/PLPB/PM* NRR/DE/EPNB* NRR/DLP/PLPB/BC NRR/DLP/PLPB/PM NAME JHolonich DAlley DMorey JHolonich DATE 10/24/2018 10/24/2018 10/25/2018 10/29/2018