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Summary of 04/12/2018 Meeting with the State of New York Regarding the Indian Point Generating Unit No. 2 Reactor Pressure Vessel Head Penetration Alternative Weld Repair
	ML18131A057
Person / Time
Site:	Indian Point
Issue date:	05/11/2018
From:	Richard Guzman; Plant Licensing Branch 1
To:	;
	Guzman R
Shared Package
ML18131A058	List: ML18131A057;
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	Download: ML18131A057 (8)
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May 11, 2018 PARTICIPANTS:

Representatives from the State of New York

SUBJECT:

SUMMARY

OF APRIL 12, 2018 MEETING WITH THE STATE OF NEW YORK REGARDING THE INDIAN POINT GENERATING UNIT NO. 2 REACTOR PRESSURE VESSEL HEAD PENETRATION ALTERNATIVE WELD REPAIR On April 12, 2018, a Category 1 public meeting was held between the U.S. Nuclear Regulatory Commission (NRC) staff and representatives from the state of New York. The purpose of the meeting was to discuss Entergy Operations, lnc.'s (Entergy; the licensee) relief request submittal to use an alternative method for addressing a reactor penetration weld flaw for Indian Point Generating Unit No. 2 (IP2). By letter dated April 6, 2018,1 New York State Energy Research and Development Authority (NYSERDA) requested the NRC staff to convene a public meeting to discuss NRC's analysis of Entergy's submittal before IP2 returns to service. The NRC responded to NYSERDA by letter dated April 16, 2018.2 The April 11, 2018, meeting notice and agenda are available in the Agencywide Documents Access and Management System (ADAMS) at Accession No. ML181018028. A list of attendees is enclosed.

The meeting began with opening remarks from the NRC, an overview of meeting logistics, and an introduction of the meeting participants. The state of New York, represented by NYSERDA and the New York State Department of Public Service (NYS DPS) also provided introductory remarks and questions. The NRC staff then presented an overview of the issue of the flaw, its safety significance, the licensee's embedded flaw repair, the requested American Society of Mechanical Engineers (ASME) code alternative, and the NRC staff's evaluation and conclusions as follows:

Overview of the Issue Boric acid was discovered on the reactor pressure vessel (RPV) penetration head #3 during a bare metal visual examination of the IP2 spring 2018 refueling outage. Subsequent to that discovery, additional inspections were conducted; one indication was identified as being contributing to the leakage on the head. The visual inspection found the presence of boric acid on the top of the head at the nozzle, indicating that leakage had occurred, and was characterized as an extremely small amount (significantly less than a quarter teaspoon). The boric acid was white in color indicating no corrosion of the head. The repair of the flaw has been completed.

1 ADAMS Accession No. ML18106A519 2 ADAMS Accession No. ML18106A542 Safety Significance In terms of the safety significance of the discovery, there is no current safety significance in that there is no corrosion of the head and no identified circumferential cracking in the nozzle material. There is no future safety significance in that the flaw is isolated from the environment with primary water stress corrosion cracking (PWSCC)-resistant repair weld material. The required inspection intervals for these nozzles will be every refueling outage and provide reasonable assurance of the structural integrity of the j-groove welds and the nozzle materials.

Flaw Repair The NRC staff indicated that the licensee has executed an embedded flaw repair, performed in accordance with an NRG-approved method. The staff also noted that this item was one of the questions posed in the April 6, 2018, letter from NYSERDA, and addressed the question by stating that the licensee's flaw repair method is indeed applicable to the Westinghouse Topical Report, WCAP-15987-P-A, "Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations," and the associated safety evaluation report. The NRC staff also noted that the repair is highly resistant to PWSCC as it is made from Alloy 52/52M material which has an extensive track record in both field service and in laboratory testing with respect to resistance to PWSCC. The repair method chosen by the licensee provides a significant reduction in dose to workers with respect to the code-required method. The flaw repair provides structural integrity for the nozzle and specifically, restores the integrity of the pressure boundary.

ASME Code Alternative requested by the licensee The NRC staff stated that the repair conducted by the licensee requires an alternative to the ASME code in that the flaw has not been removed. Typically, licensees request alternatives for this type of repair for a duration of 10 years or more (i.e., considered a very robust repair).

However, Entergy requested the alternative for one operating cycle. The NRC staff has evaluated the licensee's request with respect to a single operating cycle, and determined that the alternative provides an acceptable level of quality and safety for the period of the proposed duration and, therefore, is in compliance with the requirements of Title 10 to the Code of Federal Regulations (10 CFR) Section 50.55a(z)(1).

The NRC staff indicated that if there was a desire from the licensee to utilize this alternative repair for a period longer than one cycle, the licensee would need to submit an additional request for NRC evaluation with respect to continued operation. The licensee's alternative for use was authorized by the NRG on April 9, 2018 (via verbal authorization). In conclusion, the NRG staff has determined that the licensee has completed a repair that provides reasonable assurance of the adequate protection of public health and safety.

The NRC staff also responded to the questions posed in the April 6, 2018, letter, from NYSERDA, concerning (1) PWSGG and its relationship with various metals/materials present in and around the penetration and welds, (2) effectiveness of robotic repair, and (3) worker dose issues.

Primary Water Stress Corrosion Cracking PWSCC is a type of stress corrosion cracking that requires the following components prior to occurrence: (1) the presence of tensile stresses, (2) a specific environment, and (3) a susceptible material. In the case of all reactor pressure vessel upper heads, the nozzles have tensile stress due to the welds which secure them in position in the heads of reactor vessels.

The specific environment is the primary coolant at high temperatures; and the susceptible material to PWSCC which was used to fabricate the IP2 RPV head penetration nozzles and welds is Alloy 600/82/182. The NRC staff noted that the alloys resistant to PWSCC are Alloys 690/52/52M and that the licensee used Alloy 52/52M as the repair weld material.

Effectiveness of Robotic Repair The use of mechanized automated or robotic repair are extensively used in nuclear industry for a variety of reasons. Relative to manual welding, the use of robotic repair reduces radioactive dose obtained by workers significantly. Welding procedures are qualified in accordance with ASME Code; final acceptance criteria for automated versus manual welding are identical and are controlled by the ASME Code. It has been previously demonstrated that robotic repair has a higher consistency from one weld location to another as compared to manual welding repair.

Dose to Workers For this particular repair location (from underneath the RPV head), the area is a high radiation area; the principle of as low as reasonably achievable (ALARA) requires that exposures be minimized. The repair method chosen by the licensee minimizes the exposure of workers to radiation without a loss in quality and safety as compared to the code repair.

Follow-up comments from NYS DPS Representatives from the NYS DPS provided comments following the NRC staff's presentation.

A representative from NYS DPS also requested that the NRC include in any follow-up documentation being prepared regarding the specific issue, a statement that there was a thru-weld leak and that it crossed the reactor pressure boundary (for clarity and transparency for the public). He requested the description of the degradation mechanism via PWSCC be included in the documentation. The NRC staff acknowledged the request and indicated that the NRC's formal written safety evaluation will appropriately document the NRC staff's evaluation and findings of the licensee's proposed alternative.

Another representative from NYS DPS also stated that the questions in the April 6, letter have been addressed by the NRC staff through its presentation. She posed two additional questions concerning what has been done to confirm that similar thru-weld flaws do not exist on the remaining head penetrations, and whether additional flaws are more likely to occur with the discovery of the RPV penetration head no. 3 indication.

The NRC staff stated that all other nozzle/welds have been examined by ultrasonic testing examination to look for cracks and by ultrasonic testing leak path examination to assess whether there is any other leakage. All other nozzle penetrations have been examined and found acceptable for service. The NRC staff further described PWSCC in terms of ( 1) being the active degradation mechanism of concern for the RPV upper head penetration nozzles and associated J-groove welds; (2) flaw growth, and (3) the time required for these degradation types to develop based on a time-at-temperature model, which helps determine the frequency of inspection.

The NYS DPS representative further asked the NRC to describe how the NRC and the licensee accounted for any further flaw propagation by any other mechanisms such as fatigue (since the flaw had not actually been removed). The NRC staff responded by giving a description of the specific coverage area of the weld repair, the weld material as being highly resistant to PWSCC, and the embedded repair method which provides structural integrity for the nozzle. The NRC staff also indicated that the only other possible major growth mechanism after that point would be due to fatigue. The NRC staff noted that the WCAP-15987 describes a method of providing a fatigue analysis to determine the potential growth of a flaw due to fatigue. The licensee has submitted a basis for acceptance in their submittal in which they have bounded this analysis and found that fatigue will not cause significant growth of this material for the proposed duration.

Further clarification was provided by the NRC staff to indicate that the licensee's April 6, 2018, 3 supplement provided the licensee's basis for acceptance on fatigue growth and that the NRC's formal written safety evaluation will provide documentation of its completed review and authorization of the licensee's proposed alternative repair method.

During the public question and answer (Q&A) period, there were three individuals who communicated with the NRC staff as summarized below:

A representative from Riverkeeper, Inc. stated that during the inspections of the RPV head penetrations in 2016, another flaw was identified and was removed; however, for this flaw, it doesn't appear the licensee has been cited of any violations regarding pressure boundary leakage. Specifically, he asked (1) whether there were any regulatory consequences for this pressure boundary leakage, and (2) how we can be sure that more problems will not happen during the next 2-year operating cycle, given that the flaw developed during the current operating cycle.

The NRC staff responded by stating that the NRC has an inspection interval (or inspection frequency) designed to ensure that inspections are sufficiently frequent to ensure the structural integrity of the components and, therefore, provide for the safe operation of the plant. The inspection interval is also designed based on the NRC's understanding of crack growth which is highly documented through laboratory and field experience. Therefore, the inspection interval as indicated by this discovery at the beginning of the leakage cycle is appropriate.

The NRC staff indicated that based on the operating history at various plants, inspections have identified a low number of leakage issues. In this particular instance, the leaking was confined to the weld itself and was a very small crack. The NRC staff stated that it does acknowledge the specific event at IP2 was indeed, a leak across a pressure boundary. The NRC staff also noted that in terms of regulatory consequences and violations, that discussion is outside the scope of the licensing process review of the relief request. The NRC's assessment of licensee performance, inspections, and the functions of enforcement of NRC requirements is handled as a separate process with the Region.

A representative from Platts Nuclear Publications requested clarification on when the licensee's repair will be completed, when the NRC granted approval of the alternative, and whether there will be any follow-up inspections by the NRC prior to the planned IP2 shutdown in 2020. The NRC staff responded indicating that the licensee's repair was completed on or about April 9, 2018, and also referenced that the NRC's verbal authorization was also completed on April 9, 2018. The NRC stated that the licensee performs the ultrasonic testing examinations and inspections (e.g., running the ultrasonic probe over the weld); and the NRC reviews those inspections through its reactor oversight inspection program.

3 ADAMS Accession No. ML18098A088, "Indian Point, Unit 2, Fracture Mechanics Assessment of Embedded Flaw Repair Acceptability

A member of the public inquired the NRC staff concerning ALARA, asking what the total dose absorbed would be, should the licensee follow the Code repair versus total dose absorbed for non-code repair. The NRC staff responded by stating that while it does not have the specific dose values as an estimate for the discussion, the location of the weld area is a high radiation area which would have a significant amount of dose. In order to meet the code requirement, the worker would need to grind the weld area from underneath the head; and a significant amount of grinding would be necessary to take out the entire flaw in the J-groove weld which would be very difficult to do manually. Additionally, the NRC staff stated that a number of supplemental examinations of the weld joint would be necessary in order to verify the flaw removal was sufficient. The member of the public followed his question with a statement that while approval is being granted by the NRC staff, the NRC has no quality assurance and no specific license or qualifications to give the approval on the alternative. He further stated that the ASME members who wrote the Section 3, Code are very qualified, and therefore, it should be up to the ASME members to approve a methodology as an alternative to the ASME code.

No regulatory decisions were made during the public teleconference. There were no formal follow-up action items as a result of the meeting with the state of New York, nor from the Q&A period. To date, no public meeting feedback forms have been submitted through the NRC public meeting feedback system.

Please direct any inquiries to me at (301) 415-1030, or Richard.Guzman@nrc.gov Docket No. 50-247

Enclosure:

List of Attendees cc: Listserv Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

LIST OF ATTENDEES APRIL 12, 2018, MEETING WITH STATE OF NEW YORK REGARDING THE INDIAN POINT GENERATING UNIT NO. 2 REACTOR PRESSURE VESSEL HEAD PENETRATION ALTERNATIVE WELD REPAIR ATTENDEE CC C

AFFILIA TlON David Alley U.S. Nuclear Regulatory Commission (NRC)

Jay Collins NRC John Tsao NRC Seung Min NRC Richard Guzman NRC James Danna NRC Tara Inverso NRC Jenny Weil NRC Brian Haagensen NRC Andrew Siwy NRC Daniel Schroeder NRC Thomas Setzer NRC Mark Henrion NRC Stephen Pindale NRC Michael Modes NRC Doug Tifft NRC Diane Screnci NRC Neil Sheehan NRC Rob Krsek NRC Alyse Peterson State of New York, NYSERDA Bridget Frymire State of New York, Department of Public Service Tom Congdon State of New York, Department of Public Service John Sipos State of New York, Department of Public Service Bob Walpole Entergy Nuclear Operations, Inc. (Entergy)

Jerry Nappi Entergy Andrew Kattell Entergy Robert Dolansky Entergy Nelson Azevedo Entergy Rebecca Martin Entergy Robert Allen Entergy Dennis DelborQo Westchester County Peter Mccartt Westchester County Richard Webster Riverkeeper, Inc.

Margaret Coulter Riverkeeper, Inc.

David Lochbaum Union of Concerned Scientists Paul Blanch Energy Consultant Enclosure Michael Keegan Coalition for a Nuclear Free Great Lakes Bill Wallach None Elizabeth Gerke None Sarah Grubbs None Geri Shapiro Senator Gillibrand's Office Jordan Baugh Senator Gillibrand's Office Olivia Alvez Senator Schumer's Office Ali Biasotti Senator Schumer's Office Cory Hasson Congresswoman Lowey's Office Shira Siegel Congresswoman Lowey's Office James Ostroff Platts Nuclear Publications Hank Gross Mid Hudson News Allison Dunne WAMC NE Public Radio Taylor Capps Burson-Marsteller Tom Zambito The Journal News

Meeting Notice ML18101B028 Package ML18131A058 M

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ML18131A057 ee1nQ ummarv OFFICE NRR/DORL/LPL 1/PM NRR/DORL/LPL 1/LA NRR/DORL/LPL 1/BC NRR/DORL/LPL 1/PM NAME RGuzman I Betts MMarshall for JDanna RGuzman DATE 05/11/2018 05/11/2018 05/11/2018 05/11/2018