RS-17-036, Response to Request for Additional Information Regarding Relief Request I4R-10

Response to Request for Additional Information Regarding Relief Request I4R-10

ML17055B712
Person / Time
Site:	Byron
Issue date:	02/24/2017
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-17-036, TAC ME6071, TAC ME6073, TAC ME6074
Download: ML17055B712 (6)

Text

4300 Winfield Road Warrenville, IL 60555 Adomenow" A"""" ExeLon G 630 657 2000 Office 10 CFR 50.55a RS-17-036 February 24, 2017 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Response to Request for Additional Information Regarding Relief Request 14R-10

References:

(1) Letter from Jacob Zimmerman, (U. S. NRC) to M. J. Pacilio, (EGC),

"Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 Relief Requests 13R-09 and 13R-20 Regarding Alternative Requirements for Repair of Reactor Vessel Head Penetrations JAC Nos. ME6071, ME6073, and ME6074)," dated March 29, 2012 (M L 120790647)

(2) Letter from David M. Gullott, (EGC) to U.S. NRC, "Relief Request for Alternative Requirements for the Repair and Examination of Reactor Vessel Head Penetrations for the Fourth Inservice Inspection Interval," dated August 16, 2016 (ML16229A250)

(3) Email from Joel Wiebe, (U.S. NRC) to Jessica Krejcie (EGC),

"Preliminary Request for Additional Information (RAI) Regarding Relief Request 14R-10," dated December 8, 2016 (ML16343A252)

(4) Letter from David M. Gullott, (EGC) to U.S. NRC, "Response to Request for Additional Information for Byron Station Relief Request 14R-10: Proposed Alternative Requirements for the Repair and Examination of Reactor Vessel Head Penetrations for the Fourth Inservice Inspection Interval," dated December 29, 2016 (ML17003A274)

(5) Letter from David M. Gullott, (EGC) to U.S. NRC, "Supplemental Response to Request for Additional Information for Byron Station Relief Request 14R-10: Proposed Alternative Requirements for the

February 24, 2017 U.S. Nuclear Regulatory Commission Page 2 Repair and Examination of Reactor Vessel Head Penetrations for the Fourth Inservice Inspection Interval," dated February 13, 2017 (ML17044A294)

(6) Email from Joel Wiebe, (U.S. NRC) to Jessica Krejcie (EGC),

"Request for Additional Information (RAI) Regarding Relief Request 14R-10," dated February 21, 2017 (7) Letter from Herbert N. Berkow, (U. S. NRC) to Henry A. Sepp, (Westinghouse Electric Company), "Acceptance for Referencing Topical Report WCAP-15987-P, Revision 2, 'Technical Basis for the Embedded Flaw Repair Process for Repair of Reactor Vessel Head Penetrations,"' dated July 3, 2003 (ML031840237)

In Reference 2, Exelon Generation Company, LLC (EGC) requested the U.S. Nuclear Regulatory Commission (NRC) to authorize an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) regarding the repair of reactor vessel head penetrations (VHPs) at Byron Station Units 1 and 2 (Byron) for the fourth 10-year Inservice Inspection (ISI) interval. The request also included ISI examination requirements for future VHP repairs as well as previously repaired VHPs. This request was consistent with that previously approved by the NRC in Reference 1 for the third ISI interval at Braidwood and Byron Stations. The NRC requested additional information related to their review of Reference 2 in Reference 3. EGC submitted a response to the Reference 3 request in Reference 4. EGC supplemented the Reference 4 response in Reference 5. During a teleconference on February 21, 2017, the NRC indicated that additional information was needed to support their review. The NRC requested additional information related to their review in Reference 6. The information below provides the response to the NRC's Reference 6 request.

Specifically, Reference 6 included the following:

'By letter dated August 16, 2016, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16229A250), Exelon Generation Company, LLC (the licensee) requested the U.S. Nuclear Regulatory Commission (NRC) to authorize an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) regarding the repair of reactor vessel head penetrations (VHPs) at Byron Station Units 1 and 2 (Byron) for the fourth 10-year ISI interval. The licensee's request also included inservice inspection (ISI) examination requirements for future VHP repairs as well as previously repaired VHPs.

As discussed in a clarification call on February 21, 2017, the NRC staff needs additional information to completed its review. "

NRC RA 1 "In its February 13, 2017 letter, the licensee stated, in part, "the normal operating pressure was utilized instead of the design pressure .."

February 24, 2017 U.S. Nuclear Regulatory Commission Page 3 The NRC staff was not able to identify this change in WCAP-16401-P Revision 1 as this level of detail is not provided. Provide a discussion on the change in the pressure used in the analysis and describe the impact that this change had on the conclusions in the revision of WCAP-16401-P. Also, explain why the change in the pressure used in the analysis is appropriate and discuss how this change impacts the allowable flaw size as it relates to flaw growth as shown in Figure 3-3."

EGC Response to NRC RAI 1:

In Revision 0 of WCAP-16401 (as approved by the NRC in Reference 1), a conservative design pressure of 2500 psia was used, which resulted in an allowable flaw depth of 2.59 inches of the wall thickness. Design pressure was originally selected as a conservative, bounding input that resulted in a service life that did not challenge or limit the expected duration of flaw life due to anticipated reactor pressure vessel head (RPV) replacement. In lieu of replacing the Byron and Braidwood Stations' RPV heads, EGC has elected to peen the RPV heads as a Primary Water Stress Corrosion Cracking (PWSCC) mitigating measure. This mitigating approach resulted in the existing Embedded Flaw Repairs (EFRs) remaining for the life of the plant; necessitating a re-evaluation of the EFR service life. To support this re-evaluation, the more realistic normal operating pressure of 2250 psia was used in the WCAP-16401, Revision 1 analysis (included as Attachment 2 (Proprietary) and Attachment 3 (Non-Proprietary) of Reference 5), which is consistent with ASME Section XI requirements for evaluating operating plants, as stated in Section XI, Appendix A. In fact, the ASME Code does not specify the use of 'design pressure' with respect to flaw evaluation rules for the configuration of interest.

Primary stress limit calculations were used to calculate a maximum allowed flaw depth. The steady state normal operating pressure was used to calculate an allowable flaw depth of 3.0 inches (see Figure 1 below for schematic) based on the primary stress limit calculations.

This 3.0 inch allowable flaw depth is a uniform depth around the J-groove weld, which was used to determine the allowable service life, as discussed below. The allowable flaw depth of 3.0 inches was also validated as acceptable for all the normal and upset transient conditions, using elastic-plastic fracture mechanics methods, as specifically allowed by the Safety Evaluation of WCAP-15987-P Revision 2 (page 4 of Reference 7).

February 24, 2017 U.S. Nuclear Regulatory Commission Page 4 alodd 0.O Figure 1: Schematic of J-groove Weld The result of the primary stress limit is used as the maximum allowable flaw size (3.0 inches) for the fatigue crack growth evaluation of the postulated flaw, on the uphill or downhill sides, as it grows through the reactor vessel head thickness. The initial postulated flaw is assumed to be the entire J-groove weld depth for the governing uphill 2.54 inch side. As an illustration, Figure 2 below (which is based on Figure 3-3 of WCAP-16401-P) shows the fatigue crack growth of postulated flaws on the uphill and downhill side. Figure 2 also displays the time required for the postulated flaw to reach the maximum allowable flaw size (which was calculated based on the primary stress limits). The results from Figure 2 demonstrate that a time greater than 40 years is needed for the initial postulated flaw to reach the maximum allowable flaw size.

February 24, 2017 U.S. Nuclear Regulatory Commission Page 5 00 Pressure = 2250 psia, Allowable Flaw Size = 3.0" 280 Uphill Pressure = 2500 psia, Allowable Flaw Size = 2.59" 1 2.60 240 U

e

~a N 220 R

LL 200 1.80 Downhill 1.60 1.40 0 5 10 15 20 25 30 35 40 Period (Years)

Figure 2: Fatigue Crack Growth Prediction in the Reactor Vessel Head with Maximum Postulated Flaws in the Attachment Weld (based on Figure 3-3 of WCAP-16401-P Rev. 1 included in Reference 5 Attachments 2 and 3)

February 24, 2017 U.S. Nuclear Regulatory Commission Page 6 RIGINC f11~,~1 2.

WCAP-16401-P Revision 1 states "Engineering evaluations were performed and the results are presented in this report to provide the maximum flaw sizes that would satisfy the requirements in Section IX [XI] of the ASME Code [2004 Edition]."

Verify that all of the analyses, documented in WCAP-16401-P Revision 1, meet the ASME Code Section XI, 2007 Edition with the 2008 Addenda which is the applicable Code Edition and Addenda for the fourth ISI interval at Byron Units 1 and 2.

E GC Response to NrRC RA! 2:

The methodologies used in the evaluations contained in WCAP-16401-P Revision 1 as included in Reference 5 Attachment 2 (Proprietary) and Attachment 3 (Non-Proprietary) are consistent with the requirements of Appendix A and Appendix K of Section XI. There are no differences in the fracture mechanics evaluations and methodologies (i.e. elastic-plastic fracture mechanics guidance ASME Section XI Appendix K or fatigue crack growth ASME Section XI Appendix A) between the ASME Section XI 2004 Edition and the 2007 Edition/2008 Addenda. Therefore, the WCAP-16401-P Revision 1 results and conclusions between 2004 Edition and the 2007 Edition/2008 Addenda would be the same.

There are no regulatory commitments contained in this letter.

If you have any questions regarding this matter, please contact Jessica Krejcie at (630) 657-2816.

Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC cc: Regional Administrator- NRC Region III NRC Senior Resident Inspector- Byron Station NRC Project Manager, NRR Braidwood and Byron Station Illinois Emergency Management Agency Division of Nuclear Safety