SBK-L-15187, 3 - Responses to Request for Additional Information Related to the Review of the License Renewal Application - Set 24

3 - Responses to Request for Additional Information Related to the Review of the License Renewal Application - Set 24

ML15287A396
Person / Time
Site:	Seabrook
Issue date:	10/09/2015
From:	Dean Curtland NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-15187, TAC ME4028
Download: ML15287A396 (11)

Text

iJiE/Tiera ENERGY.

~sEABRooK October 9, 2015 10 CFR 54 SBK-L-15 187 Docket No. 5 0-443 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Seabrook Station Supplement 43 - Responses to Request for Additional Information Related To the Review of the Seabrook Station, Unit 1 License Renewal Application - Set 24 (TAC No. ME4028')

References:

1. NextEra Energy Seabrook, LLC letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License," May 25, 2010. (Accession Number ML101590099)

2. NextEra Energy Seabrook, LLC letter SBK-L-15073, "Responses Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)," May 26, 2015. (Accession Number ML15149A279)

3. NRC Request For Additional Information Related To The Review Of The Seabrook Station, Unit 1, License Renewal Application - Set 24 (TAC No. ME4028); August 28, 2015 (Accession Number ML15224A566)

4. LR-ISG-20 11-04, "Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors," May 28, 2013.

5. EPRI Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), Technical Report 1022863, December 2011.

In Reference 1, NextEra Energy Seabrook, LLC (NextEra Energy Seabrook) submitted an application for a renewed facility operating license for Seabrook Station Unit 1 in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.

NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

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U.S. Nuclear Regulatory Commission SBK-L-15187/Page 2 In Reference 2, NextEra Energy submitted a revised PWR Vessel Internals Program using the guidance provided in LR-ISG-201 1-04 (Ref. 4) and MRP-227-A (Ref. 5).

In Reference 3, the NRC requested additional information in order to complete the review of NextEra Energy Seabrook's License Renewal Application (LRA). provides NextEra Energy Seabrook' s response for the requested information.

Provided in this Supplement are changes to the LRA. To facilitate understanding, the changes are explained, and where appropriate, portions of the LRA are repeated with the change highlighted by strikethroughs for deleted text and bolded italics for inserted text. There are no new or revised regulatory conmmitments contained in this letter.

If there are any questions or additional information is needed, please contact Mr. Edward J.

Carley, Engineering Supervisor - License Renewal, at (603) 773-7957.

If you have any questions regarding this correspondence, please contact Mr. Michael H. Ossing Licensing Manager, at (603) 773-7512.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on October 9, 2015.

Sincerely, NextEra Energy Seabrook, LLC Dean Curtland Site Vice President - Supplement 43 - Responses to Request for Additional Information Related To the Review of the Seabrook Station, Unit 1 License Renewal Application cc:

D. H. Dorman NRC Region I Administrator J. G. Lamb NRC Project Manager P. C. Cataldo NRC Senior Resident Inspector R. A. Plasse NRC Project Manager, License Renewal L. M. James NRC Project Manager, License Renewal

U.S. Nuclear Regulatory Commission SBK-L-15187/Page 3 Mr. Perry Plummer Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 Mr. John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 to SBK-L-15187 Supplement 43 - Responses to Request for Additional Information Related To the Review of the Seabrook Station, Unit 1 License Renewal Application

United States Nuclear Regulatory Commission SBK-L-15 187 / Enclosure 1/ Page 2 RAI 3.0.3.3.5-1

Background

By letter dated May 26, 2015, the applicant provided its response to A/LAI 3 of the MRP-227-A report. The applicant stated that the original equipment Alloy X-750 control rod guide tube (CRGT) split pins were proactively replaced at Seabrook with cold worked Type 316 stainless steel split pins to mitigate potential stress corrosion cracking concerns based on industry operating experience. In addition, the applicant explained that there are currently no vendor specific requirements to inspect the replacement CRGT split pins; however, through the station's participation in industry groups and the evaluation of industry operating experience this position may change as warranted. The applicant stated that plants that have already installed the replacement Type 316 stainless steel split pins will provide a leading indicator for operating experience, and Seabrook will evaluate any split pin failures from these plants.

The staff noted that Section 3.2.5.3 of the SE, Rev.1, for MRP-227 states, in part, that it is recommended that the evaluation performed by the applicant in response to A/LAI 3 "consider the need to replace the Alloy X-750 support pins (split pins), if applicable, or inspect the replacement Type 316 stainless steel support pins (split pins) to ensure that cracking has been mitigated and that aging degradation is adequately monitored during the extended period of operation."

Issue Since the applicant has already replaced all of its X-750 split pins at Seabrook and the applicant is not proposing to inspect the replacement Type 316 stainless steel support pins (split pins) it is not clear to the staff how the applicant will ensure that cracking has been mitigated and that aging degradation is adequately monitored during the period of extended operation (PEO). The applicant did not provide adequate justification how split pin operating experience and inspections of CRGT split pins at other plants will be used to provide a reasonable forecast for plant-specific operating experience at Seabrook.

Req uest If inspection of the replacement Type 316 stainless steel support pins (split pins) is not proposed, provide the basis that the concerns documented in Section 3.2.5.3 of the SE, Rev.1 for MRP-227 and A/LAI 3 are adequately addressed in the LRA and that age-related degradation is adequately monitored during the PEO.

Provide the following information if operating experience and inspections of CRGT split pins at other Westinghouse-designed facilities will be used as part of the basis for managing potential cracking in the CRGT split pins at Seabrook Station: (a) identify the plants that will be performing inspections of their replaced Type 316 cold-worked CRGT

United States Nuciear Regulatory Commission SBK-L-15187 / Enclosure 1/ Page 3 split pins on behalf of Seabrook Station, (b) identify the process or processes that will be used in accordance with the "Administrative Controls" or "Confirmation Process elements of the PWR Reactor Internals Program to collect and compile the inspection data from these plants, (c) identify the criteria that will be implemented in accordance with the "monitoring and trending" program element of the AMP to assess the data from the other plants, and (d) identify the plant-specific "acceptance criteria" that will be used to assess such data and the "corrective actions" that will be taken if the acceptance criteria are not met.

NextEra Enerqy Seabrook Response to RA! RAI 3.0.3.3.5-1 The MRP-227-A, subsection 4.4.3 [3] guidance for guide tube support pins (split pins) in Westinghouse plants is limited to plant-specific recommendations. The owner is directed to review and follow the original equipment manufacturer (OEM) recommendations for aging management and subsequent performance monitoring.

Results of the detailed categorization and ranking of internals components contained in MRP-227-A, Table 3-3 [3] identify only X-750 split pins as requiring specific actions to manage material aging in the PEO; thus, no inspection or monitoring of the 316 stainless steel (SS) variant for CRGT support pins is included in MRP-227-A, Table 4-9

[3]. As indicated in the Seabrook Station, Aging Management Program Basis Document

[5], Seabrook Unit 1 followed the OEM recommendation to replace the originally installed X-750 CRGT support pins with support pins fabricated from strain-hardened 3163SS material during Refueling Outage 11 (Fall 2006).

As listed in MRP-191, Table 5-1 [6], the 316 SS guide tube support pins were screened-in for the aging degradation mechanisms of wear, fatigue, and irradiation stress relaxation/irradiation creep (ISR/IC). In MRP-232 [7], the 316 33 CRGT support pins were categorized as MRP-1 91 Category A, "no additional measures." The OEM recommendations do not require subsequent inspection of the 316 SS support pins.

Long-term material behavior has been extensively studied, and from past testing and field experience, all identified degradation mechanisms, including irradiation assisted stress corrosion cracking (IASOC), stress corrosion cracking (SCC), wear, fatigue, ISR/IC, and embrittlement, have been assessed, and it was concluded that the 3163SS CRGT support pins will perform all intended functions [4]. Therefore, there is no requirement for augmented inspections of the Seabrook Unit 1, 316 SS support pins [3 and 5]. Seabrook Unit I existing programs comply with the OEM recommendations and MRP-227-A adequacy evaluation requirements for aging management of reactor internals CRGT 316 33 support pins. Appropriate actions would be taken upon receiving further recommendations from Westinghouse Electric Company LLC or as a result of operating experience.

As shown in response to RAI 3.0.3.3.5-2 below NextEra Seabrook has enhanced B.2.1.7 PWR Vessel Internals; Element 10 Operating Experience to confirm participation in PWROG activities and appropriate action upon receiving further

U'nited States Nuclear Regulatory Commission SBK-L-15187 / Enclosure 1/ Page 4 recommendations from Westinghouse Electric Company LLC or as a result of operating experience.

RAI 3.0.3.3.5-2 Backgqround:

LRA Table 3.1.2-3, Reactor Vessel Internals, as amended by letter dated May 26, 2015, indicates that the clevis insert bolts are nickel alloy and that loss of material will be managed by the PWR Vessel Internals Program. In addition, the staff noted that, Table 1, of the May 26, 2015, letter indicates that the clevis insert bolts are managed as part of inspections performed in accordance with ASME Code Section XI.

Appendix A to MRP-227-A indicates that failures of Alloy X-750, precipitation-hardenable nickel-chromium alloy clevis insert bolts were reported by one Westinghouse designed plant in 2010. Furthermore, the staff noted that these clevis insert bolts failed because of cracking, which is an aging effect that was not addressed in MRP-227-A.

Issue:

It is not clear to the staff whether the generic operating experience associated with Westinghouse-design clevis insert bolt cracking is applicable to the design of the clevis insert assemblies and clevis insert bolts at Seabrook Station and, if so, how the PWR Vessel Internals Program will be used to monitor and manage cracking of clevis insert bolts at the plant, including the potential need for adjusting the "Existing Program"s protocols for clevis insert bolts in the MRP-227-A report.

Request:

Discuss and justify whether the generic operating experience associated with cracking of the Westinghouse-design clevis insert bolts is applicable to design of the clevis insert assemblies and clevis insert bolts at Seabrook Station.

1. If it is not applicable to the design of the clevis insert assemblies at Seabrook, explain and justify why the operating experience does not apply to the plant.

2. If the generic operating experience is applicable to design of the clevis insert assemblies and clevis insert bolts at Seabrook Station, explain and provide a basis for each of the following items:

(a) Specify the material of fabrication, including any applicable heat treatment, for the clevis insert bolts that were procured for the station and are currently in-service.

Clarify (explain) whether the material of fabrication is susceptible to:

(i)fatigue induced cracking, or

iUnited States Nuclear Regulatory Commission SBK-L-15187 / Enclosure 1/ Page 5.

(ii) cracking induced by any of the stress corrosion cracking, or (iii) intergranular attack mechanisms.

(b) Clarify how the PWR Vessel Internals program or an alternative AMP will be used to monitor and manage cracking of clevis insert bolts at the plant:

(i) If the PWR Vessel Internals Program will be used for aging management of the clevis insert bolts, clarify how the "Existing Program" protocols of the AMP will be augmented (if at all) to inspect the clevis insert bolts in a manner that is capable of detecting cracking in the bolts.

(ii) If the PWR Vessel Internals Program will not be augmented, provide a technical justification for the adequacy of the existing VT-3 visual inspection method to detect cracking in the bolts before a loss of clevis insert assembly intended function.

(iii) If an alternative AMP will be used for aging management of cracking in the bolts, identify the AMP and clarify the "Detection of Aging Effects" and "Monitoring and Trending" program element criteria that will be used to monitor for cracking. The response should include details about the method and frequency that will be used for the inspections and sample of clevis insert assemblies and bolts that will be inspected out of the total population of clevis insert' assemblies and bolts that are included in the plant design.

NextEra Energyv Seabrook Response to RAI 3.0.3.3.5-2 The generic operating experience is applicable to the design of the clevis insert assemblies and clevis insert bolts at Seabrook Station. Operating experience (GE) related to clevis insert cap screw (bolt) degradation was reported in [8] where during a 10-year in-service inspection in 2010, at another plant, there was visual evidence of degradation of clevis insert bolt heads and dowel pins. A root cause analysis [11]

identified the cause of the clevis insert cap screws failures at the reference plant to be primary-water stress corrosion cracking (PWSCC) due to the use of Inconel X-750 with a susceptible heat treatment.

The material of fabrication for the clevis insert bolts that were procured for Seabrook Unit I is Alloy X-750. As noted in [8], all Westinghouse-designed plants were confirmed to contain similar heat treatment of Alloy X-750 as that used for the subject cap screws.

Seabrook was included in the plants considered in the fleet and thus, the generic operating experience associated with cracking of the Westinghouse-design clevis insert bolts is applicable to Seabrook Station.

U'nited States Nuclear Regulatory Commission SBK-L-15 187 / Enclosure 1/ Page 6 Since the Seabrook design is applicable to the generic design, susceptibility of the X-750 material to the degradation mechanisms is based on MRP evaluations. The inclusion of the clevis insert bolts in MRP-227-A [3] was based on the screening and expert panel evaluation documented in MRP-1 91 [6], which determined that the bolts could be susceptible to stress corrosion cracking (SCC) and wear. As stated in MRP-175 [12], the degradation mechanism of SOC includes intergranular SCC, transgranular SCC, PWSCC, and low-temperature crack propagation. These mechanisms are also evaluated in [11] and the root cause of the clevis insert cap screws failures at the reference plant is identified as PWSCC.

Based on the MRP, screening (SCC and wear), categorization and ranking of the reactor internals components, the clevis insert bolts were categorized as an Existing component

[3, Table 4-9] where the ASME Section Xl inspection is credited for aging management of the clevis.inserts. The clevis insert bolt was screened for mechanisms that would exhibit cracking, but the VTI-3 inspection is intended to detect wear, since wear of the clevis/insert is the issue identified for active management under MRP-227-A [3]. It was determined that the ASME Section Xl inspection was an adequate existing program to monitor the clevis insert bolts for aging effects [3].

The PWR vessel internals program will be used for aging management of the Seabrook Unit 1 clevis insert bolts. The potential aging degradation of the clevis insert bolts at Seabrook Unit 1 will be managed according to the NRC-approved topical report, MRP-227-A. Thus, the clevis insert bolts are listed in [5, Table 1] as an Existing Programs component. The Seabrook Unit 1 clevis insert assembly is included in the Seabrook ASME Section Xl program. [The clevis inserts were inspected as part of the Seabrook 10-year in-service inspection (ISI) in OR 13 (2009.) using a VT-3 inspection [9 and 10].

None of these inspections have returned relevant indications for the clevis insert bolts.

The OE related to the failure of the Alloy X-750 clevis insert bolts is discussed in [3, Appendix A]. The clevis bolt failures lead to additional guidance, for the scope of the Existing (ASME Section Xl, VT3) inspection, which was provided in Technical Bulletin TB-14-5 [8]. This guidance did not change the type or timing of the inspection but gave recommendations for the scope and focus of the examination in order to detect known indications of failure.

Seabrook is participating in the PWR Owners Group (PWROG) program, PA-MSC-1198. This program is intended to address the lower radial support system designs across the PWR fleet in providing a technical basis that supports the current aging management strategy for the clevis insert bolts, considering the clevis failure OE.

Future inspections of the Seabrook Unit 1 clevis insert assembly will consider recommendations provided in Westinghouse technical bulletin TB-i14-5 [8] and, as applicable, any additional recommendations from the PWROG Program, PA-MSC-1 198.

United States Nuclear Regulatory Commission SBK-L-15 187 / Enclosure 1/ Page 7 In response to RAI 3.0.3.3.5-2, the following changes have been made to the LRA:

The fourth paragraph of B.2. 1.7 PWR Vessel Internals; Element 10 Operating Experience is revised as follows:

FPLNextEra participates in the industry programs for investigating and managing aging effects on PWR Vessel Internals including EPRI and the Pressurized Water Reactor Owners Group (PWROG). Through its participation in EPRI MRP activities, F-P-L NextEra and Seabrook Station will continue to benefit from the reporting of PWR Vessel Internals inspection information, and will share its own internals inspection results with the industry, as appropriate. The Seabrook Station PWR Vessel Internals Program will implement applicable results of the industry programs. Seabrook will implement appropriate actions upon receiving further recommendations from Westinghouse Electric Company LLC or as a result of operating experience.

Future inspections of the Seabrook Unit I clevis insert assembly will consider recommendations provided in Westinghouse technical bulletin TB-14-5 and, as applicable, any additional recommendations from the PWROG Program, PA-MSC-1 198.

References

1. U.S. Nuclear Regulatory Commission Letter, "Request for Additional Information Related to the Review of the Seabrook Station Unit 1 License Renewal Application -

Set 24 (TAC No. ME4028)," August 28, 2015. (ADAMS Accession No. ML15224A566)

2. NextEra Energy Seabrook Letter, SBK-L-1 5073, "Responses to Applicant/Licensee Action Items for the Inspection and Evaluation Guidelines for Pressurized Water Reactor Vessel Internals (MRP-227-A)," May 26, 2015. (ADAMS Accession No. ML15149A279)

3. Materials Re/liab ility Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

4. Westinghouse Report, WCAP-16557-P, Rev. 1, "FPL Energy LLC, Seabrook Unit 1 Replacement CW 316 Support Pin Design Equivalency Report," May 2006 (Proprietary).

5. Seabrook Station License Renewal Project Document, LRAP-M016, Rev. 4, "Aging Management Program Basis Document, PWR Vessel Internals," May 22, 2015.

6. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-19 1). EPRI, Palo Alto, CA: 2006. 1013234.

7. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232, Revision 1). EPRI, Palo Alto, CA: 2012. 1021029.

' L0nited States Nuclear Regulatory Commission SBK-L-15187 / Enclosure 1/ Page 8

8. Westinghouse Technical Bulletin, TB-14-5, "Reactor Internals Lower Radial Support Clevis Insert Cap Screw Degradation," August 25, 2014.

9. Seabrook Station Engineering Procedure, ES1807.007, Rev. 05, "Reactor Vessel Interior Visual Inspection Procedure."

10. Seabrook Station Procedure, ES1807.025, Rev. 06, "In-service Inspection (ISl)

Visual Examination Procedure."

11. CNP Document, Enclosure 2to AEP-NRC-2014-59, I&M CAP Document AR 1010-1804-10, Root Cause Evaluation Attachment, "Rx Vessel core Support Lug Bolting Anomalies," (ADAMS Accession No. ML14253A317)

12. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-1 75). EPRI, Palo Alto, CA:

2005. 1012081.