ML20345A008

From kanterella
Jump to navigation Jump to search
Summary of Audit of EPRI for SGTF Information Related to TSTF-577
ML20345A008
Person / Time
Issue date: 12/15/2020
From: Paul Klein
NRC/NRR/DNRL/NCSG
To: Leslie Terry
NRC/NRR/DNRL/NCSG
Terry L, 301-415-1167
Shared Package
ML20216A673 List:
References
EPID L-2020-PMP-0005
Download: ML20345A008 (8)
v • d • e


Text

December 15, 2020 MEMORANDUM TO: Steven D. Bloom, Chief Corrosion and Steam Generator Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation FROM: Paul A. Klein, Senior Materials Engineer /RA/

Corrosion and Steam Generator Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation

SUBJECT:

AUDIT

SUMMARY

FOR THE REGULATORY AUDIT OF ELECTRIC POWER RESEARCH INSTITUTE FOR STEAM GENERATOR TASK FORCE INFORMATION RELATED TO TECHNICAL SPECIFICATION TASK FORCE-577, REVISED FREQUENCIES FOR STEAM GENERATOR TUBE INSPECTIONS (EPID: L-2020-PMP-0005)

The U.S. Nuclear Regulatory Commission (NRC) staff conducted an audit of the Electric Power Research Institute from August 17, 2020 - September 24, 2020, to review Steam Generator Task Force information related to Technical Specification Task Force-577, Revision 0, Revised Frequencies for Steam Generator Tube Inspections, which was submitted to the NRC for review on June 8, 2020 (Agencywide Documents Access and Management System Accession No. ML20160A359). Specifically, the Audit Team reviewed information to gain a better understanding of information underlying the application to evaluate the technical basis supporting the inspection interval extension proposed for thermally treated Alloy 600 steam generator tubing. is the audit summary that describes the content of the review, identifies audit participants, and documents audit observations. Enclosure 2 identifies audit entrance and exit meeting participants.

Enclosures:

1. Audit Summary
2. Meeting Participants CONTACT: Paul Klein, NRR/DNRL 301-415-4030

Pkg: ML20216A673 Memo: ML20345A008

  • via e-mail NRR-106 OFFICE NRR/DNRL/NCSG NRR/DNRL/NRLB/LA NRR/DNRL/NCSG/BC NAME PKlein SGreen* SBloom*

DATE 12/09/2020 12/10/2020 12/15/2020 U.S. NUCLEAR REGULATORY COMMISSION AUDIT

SUMMARY

FOR THE REGULATORY AUDIT OF ELECTRIC POWER RESEARCH INSTITUTE FOR STEAM GENERATOR TASK FORCE INFORMATION RELATED TO TECHNICAL SPECIFICATION TASK FORCE-577, REVISED FREQUENCIES FOR STEAM GENERATOR TUBE INSPECTIONS I. Background During a public meeting with the industry Steam Generator Task Force (SGTF) on February 13, 2019, industry representatives informed the U.S. Nuclear Regulatory Commission (NRC) staff that they planned to submit a new Technical Specification Task Force (TSTF) Traveler to revise the steam generator (SG) inspection intervals (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML19044A416).

On September 10, 2019, the TSTF submitted a draft of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, that proposed extending the maximum interval between SG inspections (ADAMS Accession No. ML19254B397). On September 18, 2019, the TSTF submitted a revised draft TSTF-577 that proposed basing the SG inspection frequency on an operational assessment (OA) and removing the prescriptive limit on the maximum interval between inspections (ADAMS Accession No. ML19301A001). The NRC staff held information exchanges with the TSTF and industry SGTF on draft TSTF-577 during public meetings held on January 22, 2020 (Package ADAMS Accession No. ML20041E013), and February 24, 2020 (Package ADAMS Accession No. ML20066E421). During the February 24, 2020, public meeting, the industry SGTF presented information on an OA feasibility study for thermally treated Alloy 600 (Alloy 600TT) to support extending the inspection interval for Alloy 600TT SG tubing.

On June 8, 2020, the TSTF submitted TSTF-577, Revision 0 (ADAMS Accession No. ML20160A359) to the NRC for review. Based on the information presented in the February 24, 2020, public meeting, and the information contained in the TSTF-577 license amendment request (LAR), the NRC staff requested an audit of the technical basis documents supporting the TSTF-577 LAR for Alloy 600TT SG tubing.

The purpose of the audit was to: (1) gain a better understanding of information to evaluate the technical basis supporting the inspection interval extension proposed for Alloy 600TT SG tubing, and (2) identify information that will require docketing to support the basis of the regulatory decision. Specifically, the Audit Team reviewed information related to the Alloy 600TT operational feasibility study. The Audit Plan for the Regulatory Audit of Electric Power Research Institute for Steam Generator Task Force Information Related to Technical Specification Task Force-577, Revised Frequencies for Steam Generator Tube Inspections, is available in ADAMS under Accession No. ML20216A676.

II. Regulatory Audit Bases An audit was required to examine detailed information related to extending the inspection interval for Alloy 600TT SG tubing and reach a safety conclusion on TSTF-577. The NRC staff must have sufficient information to ensure that acceptable risk and reasonable assurance of safety can be documented in the NRC staffs safety evaluation.

Enclosure 1

The regulatory audit was based on the following regulations:

  • General Design Criteria (GDC) 14, Reactor Coolant Pressure Boundary, in Appendix A, General Design Criteria for Nuclear Power Plants, of Part 50, Domestic Licensing of Production and Utilization Facilities, in Title 10 of the Code of Federal Regulations (10 CFR), Energy, requires that the reactor coolant pressure boundary (RCPB) shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
  • GDC 30, Quality of Reactor Coolant Pressure Boundary, in Appendix A of 10 CFR Part 50, requires, in part, that components which are part of the RCPB shall be designed, fabricated, erected, and tested to the highest quality standards practical.
  • GDC 31, Fracture Prevention of Reactor Coolant Pressure Boundary, in Appendix A of 10 CFR Part 50, requires, in part, that the RCPB shall be designed with sufficient margin to ensure that - when stressed under operating, maintenance, testing, and postulated accident conditions - the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is minimized.
  • GDC 32, Inspection of Reactor Coolant Pressure Boundary, in Appendix A of 10 CFR Part 50, requires, in part, that the RCPB be designed to permit periodic inspection and testing to assess structural and leakage integrity.
  • Paragraph (g) in 10 CFR 50.55a, Codes and Standards, requires that in-service inspection programs meet the applicable inspection requirements in Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
  • 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, requires that licensees be able to monitor the condition of the SG tubes to provide reasonable assurance that the tubes are capable of fulfilling their intended functions.
  • Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, of 10 CFR Part 50 applies to implementation of the Steam Generator Program.

III. Audit Location and Dates Dates: August 17, 2020 - September 24, 2020 Location: Remote Online Audit Audit teleconferences were held on August 17, 19, and 27; and September 1, 3, 8, 10, 15, 17, and 24, 2020.

IV. Audit Team Members Paul Klein (NRR/DNRL/NCSG, Senior Materials Engineer, Audit Lead)

Leslie Terry (NRR/DNRL/NCSG, Materials Engineer)

Andrew Johnson (NRR/DNRL/NCSG, Materials Engineer)

Gregory Makar (NRR/DNRL/NCSG, Materials Engineer)

Steven Bloom (NRR/DNRL/NCSG, Branch Chief)

Pat Purtscher (RES/DE/CMB, Materials Engineer)

Sasan Bakhtiari (Argonne National Laboratory, NRC Contractor)

V. Applicant and Industry Staff Participants Helen Cothron (Steam Generator Management Program)

Various members of industry SGTF Supporting personnel See Enclosure 2 for a detailed list of audit entrance and exit meeting participants VI. Documents Audited In addition to the documents listed below, the NRC staff reviewed presentation materials provided for the audit teleconferences and reviewed eddy current inspection data for Alloy 600TT SG tubing.

A. Westinghouse, SG-CDMP-20-16-NP, Revision 0, A Study in Operational Assessment for Thermally Treated Alloy 600 Steam Generator Tubing, July 2020.

B. Intertek, AIM-190610636-2-2, Feasibility Study for the Potential to Extend Inspection Intervals for the A600TT Fleet, June 30, 2020.

VII. Summary of Observations Westinghouse and Intertek performed feasibility studies to evaluate and document the potential for extending the operating period between eddy current inspections for U.S.

plants with Alloy 600TT SG tubing, beyond the current TS limits.

The Audit Team was provided draft copies of the documents identified in Section VI of this audit summary prior to the audit, reviewed presentation materials prepared for the audit meetings identified in Section III of this audit summary, and reviewed eddy current inspection data for select Alloy 600TT SG tubing with cracks during two of the remote audit sessions.

As part of the feasibility studies, Westinghouse collected raw eddy current data for all stress corrosion cracking (SCC) indications reported in U.S. plants with Alloy 600TT SG tubing.

The eddy current data was reanalyzed by a single data analyst to obtain a consistent evaluation approach, including indication detection and sizing. The reanalysis also looked at the eddy current data for the prior two inspections for each SCC indication to determine if a precursor or missed indication was present. Based on reanalysis, a total of nine indications were determined to be either false calls or benign signals and were not included in the crack growth calculations. Westinghouse calculated crack growth rates from the reanalyzed eddy current data. Intertek did not have access to the raw eddy current data, therefore, to perform the feasibility studies, they collected all readily available SCC indication information for the U.S. plants with Alloy 600TT SG tubing. Intertek did not reanalyze the eddy current data.

Excluding tube end cracking, nine of the seventeen U.S. plants with Alloy 600TT SG tubing have reported SCC, including axial and circumferential outside diameter SCC (ODSCC) and axial and circumferential primary water SCC (PWSCC). There has been a total of 205 SCC indications in 135 tubes reported. Tube locations where SCC has been reported include the top of tubesheet expansion transition, tubesheet expansion zone, tube support plate (TSP), low row U-bend, freespan ding, TSP dent, and freespan. Using the Alloy 600 TT fleet data, Westinghouse and Intertek performed full bundle fully probabilistic OAs to determine the feasibility of extending the time between inspections.

Westinghouse focused their OA feasibility study on two degradation mechanisms; PWSCC and circumferential ODSCC at tubesheet expansion transitions. Westinghouse performed a baseline probabilistic analysis with sensitivity studies to determine how inputs such as crack growth rates, tube noise, probability of detection (POD), and the number of flaws at the beginning of cycle would affect the results of an OA. The Westinghouse sensitivity analyses for number of cycles meeting the performance criteria were different when the Electric Power Research Institute (EPRI) default crack growth rates were used compared to the Alloy 600TT crack growth rates from eddy current reanalysis.

The Intertek feasibility study reviewed the Alloy 600TT fleet degradation history, estimated the SCC growth rates, developed SCC initiation characteristics as a function of degradation mechanism, and prepared example OAs for operating periods beyond the current TS. The Intertek OA model development continued during the course of the approximately five-week NRC staff audit. For example, additional degradation locations were considered and conclusions concerning operating intervals for some types of plants were re-evaluated relative to the initial feasibility study.

Several of the questions raised by the NRC staff during the audit included identification of high stress tubes, the probability of crack detection, and uncertainties associated with sizing cracks in the crack database. Based on staff questions, Westinghouse determined that there was an error identifying potentially higher residual stress tubes at one plant. The data for this plant and others is being re-analyzed to determine if additional tubes should be identified as potentially having higher stress. The staff asked how the assumed POD for various crack locations in the OA methodologies compared to the field experience as reflected in the fleet wide crack database. In addition, it was not clear to the staff how the uncertainty associated with growth rates determined from eddy current sizing is being considered.

VIII. Exit Briefing The NRC staff conducted an audit exit meeting on September 24, 2020. At the exit meeting the staff reiterated the purpose of the audit and stated that the audit assisted them in obtaining a better understanding of the technical basis supporting the inspection interval extension proposed for Alloy 600TT SG tubing. The staff identified information that requires docketing to support the basis of the regulatory decision. This information includes the documents identified in Section VI of this audit summary. At the exit meeting, EPRI noted that these documents are being updated based on additional analysis plus feedback from the staff during the audit and will be combined into an EPRI report that will be issued by the end of 2020 (EPRI submitted this non-public report to the NRC in a letter dated November 18, 2020 (ADAMS Accession No. ML20335A173)). The staff indicated that they would spend additional time after the audit reviewing and internally discussing the substantial amount of technical information contained in the preliminary feasibility reports. In addition, the staff discussed plans for future interactions with the TSTF on TSTF-577 and with the SGTF on developing a TS reporting requirements template, discussing the POD for SG tube inspection techniques, and revising the EPRI SG Examination and Integrity Assessment Guidelines.

IX. Requests for Additional Information Resulting from the Audit No requests for additional information were made during the audit exit meeting on September 24, 2020. However, during the audit exit meeting, the staff noted that once the final documents are received from EPRI, the staff would review the material and could have additional questions.

X. Deviations from the Audit Plan None.

REGULATORY AUDIT ENTRANCE AND EXIT MEETING PARTICIPANTS Audit Entrance Meeting Participants Audit Exit Meeting Participants Name Organization Name Organization Steven Bloom NRC Steven Bloom NRC Paul Klein NRC Paul Klein NRC Andrew Johnson NRC Andrew Johnson NRC Gregory Makar NRC Gregory Makar NRC Leslie Terry NRC Leslie Terry NRC Allen Hiser NRC Allen Hiser NRC Pat Purtscher NRC Pat Purtscher NRC Sasan Bakhtiari ANL Sasan Bakhtiari ANL William Cullen EPRI William Cullen EPRI Brian Mann Excel Services Brian Mann Excel Services Gary Whiteman Westinghouse Gary Whiteman Westinghouse Greg Glenn Westinghouse Helen Cothron EPRI Helen Cothron EPRI Jay Smith Westinghouse Jay Smith Westinghouse Jesse Baron TVA Jesse Baron TVA Lee Friant Exelon James Benson EPRI Russ Cipolla Intertek Lee Friant Exelon Scott Redner Xcel Energy Russ Cipolla Intertek Sean Kil EPRI Scott Redner Xcel Energy Kester Thompson NextEra Sean Kil EPRI Jeffrey Lanum Entergy Kester Thompson NextEra Damian Testa Westinghouse Jeffrey Lanum Entergy Kent Colgan Framatome Damian Testa Westinghouse Daniel Mayes Duke Energy Ashley Birdett Comanche Peak Steve Brown Entergy Kent Colgan Framatome Jim Skirpan Westinghouse Daniel Mayes Duke Energy Dan Folsom TVA Steve Fluit BWXT Mike Bradley Westinghouse Cotasha Blackburn TVA Enclosure 2

Retrieved from "https://kanterella.com/w/index.php?title=ML20345A008&oldid=3238625"