ML18109A116

Relief Requests Associated with the Fifth Inservice Inspection Interval

ML18109A116
Person / Time
Site:	Peach Bottom
Issue date:	04/19/2018
From:	Jim Barstow Exelon Generation Co
To:	Advisory Committee on the Medical Uses of Isotopes, Document Control Desk
References
Download: ML18109A116 (58)

Text

{{#Wiki_filter:Exelon Generation@ April 19, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 1 O CFR 50.55a Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Relief Requests Associated with the Fifth lnservice Inspection Interval Attached for your review are relief requests associated with the fifth lnservice Inspection (ISi) interval for the Peach Bottom Atomic Power Station, Units 2 and 3. The fifth interval program complies with the 2013 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. The fifth ISi interval begins on January 1, 2019 and is currently scheduled to end December 31, 2028. We request your approval of this package by April 19, 2019. There are no regulatory commitments in this letter. If you have any questions concerning this letter, please contact Tom Loomis at (61 O) 765-5510. Respectfully, Ja~=WM;: Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Relief Requests cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPS USNRC Project Manager, PBAPS R. R. Janati, Pennsylvania Bureau of Radiation Protection S. T. Gray, State of Maryland

Attachment Relief Requests ISR-02 ISR-03 ISR-04 ISR-05 ISR-06

10 CFR 50.55a Relief Request ISR-02 Revision 0 (Page 1 of 4) Request for Relief ISR-02 for Examination of Inaccessible Surfaces in Accordance with 10 CFR 50.55a(z)(2)

1.

ASME Code Component(s) Affected Code Class:

Reference:

Examination Category: Item Number:

== Description:== Component Number: Drawing Number: MC IWE-1232 E-A, E-G El.11, ES.IO Alternative Examination Requirements of ASME Section XI, Paragraph IWE-1232, "Inaccessible Surface Areas" Penetration N-3 6280-S-188 (see ML082200279)

2.

Applicable Code Edition and Addenda

The fifth ten-year interval of the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Inservice Inspection (ISI) Program (third ten-year Containment Inservice Inspection (CISI) interval) is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2013 Edition.

3.

Applicable Code Requirement

Paragraph IWE-1232(a) ("Inaccessible Surface Areas") states that portions of Class MC containment vessels, parts, and appurtenances that are embedded in concrete or otherwise made inaccessible during construction of the vessel or as a result of vessel repair/replacement activities are exempted from examination, provided: ( 1) no openings or penetrations are embedded in the concrete; (2) all welded joints that are inaccessible for examination are double butt welded and are fully radiographed and, prior to being covered, are tested for leak tightness using a gas medium test, such as Halide Leak Detector Test; and (3) the vessel is leak rate tested after completion of construction or repair/replacement activities to the leak rate requirements of the Design Specifications.

4.

Reason for Request

10 CFR 50.55a Relief Request 15R-02 Revision 0 (Page 2 of 4) In accordance with 10CFR50.55a(z)(2), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, as code requirements would require extensive structural modifications to the containment structure. When the drywell was being constructed, a 24-inch manhole was placed in the bottom head of the drywell. During construction, when the manhole was no longer needed, the penetration was seal welded, inspected, and embedded in concrete. Based on original constructions drawings, the manhole is a bolted, gasket connection. The manhole was pneumatically tested and then seal welded. The N-3 manhole was seal welded and cannot meet the paragraph IWE-1232(a)(2) code requirement for a double butt weld and fully radiographed. See Figure I5R-02-1 for more details. Adding a double butt weld would involve a modification to the drywell that would require excavation of the concrete around the bottom head of the drywell or removal of the drywell floor thus making compliance with the current code requirements a hardship without a compensating increase in the level of quality and safety.

5.

Proposed Alternative and Basis for Use PBAPS, Units 2 and 3 will exempt the Penetration N-3 components and attachments from examination, and will instead perform an Integrated Leak Rate Test (JLRT) in accordance with the PBAPS Appendix J Program, which is maintained and controlled independent of the ASME Section XI Program. The results of the most recent Type A test performed on the Primary Containment at PBAPS, Units 2 and 3 are as follows: ILRT Total Plant ILRTDate Calculation Method Weight Percent Per Day {%/day) Unit 2 November 2014 ANSI/ ANS 56.8 - 1994 0.2372 Unit 3 October 2005 Total Time Analysis based 0.2781 on BN-TOP-1, Rev. 1, 1972

6.

Duration of Proposed Alternative Relief is requested for the fifth ISI interval (third CISI interval), as well as the remaining term of the PBAPS, Units 2 and 3, renewed facility operating licenses, which currently expire on August 8, 2033, and on July 2, 2034, respectively. The remaining term of the renewed facility operating licenses refers to the PBAPS, Units 2 and 3 current fifth and upcoming sixth 120-month ISI Program intervals.

7.

Precedents 10 CFR 50.55a Relief Request ISR-02 Revision 0 (Page 3 of 4) PBAPS, Units 2 and 3, fourth ISi interval (Second CISI Interval) Relief Request 14R-48 was authorized by NRC Safety Evaluation dated February 26, 2009 (ADAMS Accession No. ML090430052). Relief Request 15R-02 for the PBAPS, Units 2 and 3, fifth ten-year ISi interval (third CISI interval), utilizes a similar approach to the previously approved relief request.

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10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 1 of 17) Request for Relief ISR-03 for Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection in Accordance with 10 CFR 50.SSa(z)(l)

1.

ASME Code Component(s) Affected Code Class: 1

Reference:

Examination Category: IWB-2500, Table IWB-2500-1 B-N-1 and B-N-2 Item Number:

== Description:== B13.10, Bl3.20, B13.30, and B13.40 Use of BWRVIP Guidelines in Lieu of Specific ASME Section XI Requirements on the Reactor Pressure Vessel Internals and Components Inspection Component Number: Vessel Interior, Interior Attachments within Beltline Region, Interior Attachments beyond Beltline Region, and Core Support Structure

2.

Applicable Code Edition and Addenda

The fifth 10-year interval of the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Inservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2013 Edition.

3.

Applicable Code Requirement

ASME Section XI requires the examination of components within the reactor pressure vessel. These examinations are included in Table IWB-2500-1 Examination Categories B-N-1 and B-N-2 and identified with the following item numbers: B13.10 B13.20 B13.30 B13.40 Examine accessible areas of the reactor vessel interior each period by the VT-3 visual examination method (B-N-1). Examine interior attachment welds within the beltline region each interval by the VT-1 visual examination method (B-N-2). Examine interior attachment welds beyond the beltline region each interval by the VT-3 visual examination method (B-N-2). Examine surfaces of the core support structure each interval by the VT-3 visual examination method (B-N-2). These examinations are performed to assess the structural integrity of the reactor vessel interior, its welded attachments, and the core support structure within the Boiling Water Reactor (BWR) pressure vessel.

10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 2 of 17) The components/welds listed in Table 2 are subject to this request for alternative. Table 2 provides only an overview of the requirements. For more details, refer to ASME Section XI, Table IWB-2500-1 and the appropriate Boiling Water Reactor Vessel and Internals Project (BWRVIP) document.

4.

Reason for Request

In accordance with 10CFR50.55a(z)(l), relief is requested for the proposed alternative to ASME Section XI requirements provided above on the basis that the use of the BWRVIP guidelines discussed below will provide an acceptable level of quality and safety. The BWRVIP Inspection and Evaluation (I&E) guidelines recommend specific inspections by BWR owners to identify material degradation with BWR components. A wealth of inspection data has been gathered during these inspections across the BWR industry. The BWRVIP guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying known or potential degradation mechanisms, and require re-examination at appropriate intervals. The scope of the BWRVIP guidelines meet or exceed that of ASME Section XI and in many instances, include components that are not part of the ASME Section XI jurisdiction. As an alternative to ASME Section XI requirements, use of BWRVIP guidelines will avoid duplicate or unnecessary inspections, while conserving radiological dose.

5.

Proposed Alternative and Basis for Use In lieu of the requirements of ASME Section XI, the proposed alternative is detailed in Table 2 for Peach Bottom Atomic Power Station, Units 2 and 3 for Examination Category B-N-1 and B-N-2. Peach Bottom Atomic Power Station, Units 2 and 3 will satisfy the Examination Category B-N-1 and B-N-2 requirements as described in Table 2 in accordance with the latest Nuclear Regulatory Commission (NRC) approved BWRVIP guideline requirements. This relief request proposes to utilize the identified BWRVIP guidelines in lieu of the associated ASME Section XI requirements, including the examination method, examination volume, frequency, training, successive and additional examinations, flaw evaluations, and reporting. Not all the components addressed by these guidelines are ASME Section XI components. The following BWRVIP guidelines are applicable to this relief request: - BWRVIP-03, Revision 19, BWR [Boiling Water Reactor] Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines - BWRVIP-06, Revision 1-A, BWR Vessel and Internals Project, Safety Assessment of BWR Reactor Internals

10 CFR 50.SSa Relief Request ISR-03 Revision 0 (Page 3 of 17) - BWRVIP-18, Revision 2-A, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines - BWRVIP-25, BWR Core Plate Inspection and Flaw Evaluation Guidelines - BWRVIP-26-A, BWR Top Guide Inspection and Flaw Evaluation Guidelines - BWRVIP-38, BWR Shroud Support Inspection and Flaw Evaluation Guidelines - BWRVIP-41, Revision 3, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines - BWRVIP-47-A, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines - BWRVIP-48-A, Vessel ID [Internal Diameter] Attachment Weld Inspection and Flaw Evaluation Guidelines - BWRVIP-49-A, Instrument Penetration Inspection and Flaw Evaluation Guidelines - BWRVIP-76, Revision 1-A, BWR Core Shroud Inspection and Flaw Evaluation Guidelines - BWRVIP-94NP, Revision 2, Program Implementation Guide BWRVIP-138, Revision 1-A, Updated Jet Pump Beam Inspection and Flaw Evaluation - BWRVIP-180, Access Hole Cover Inspection and Flaw Evaluation Guidelines - BWRVIP-183-A, Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines Inspection services, by an Authorized Inspection Agency, will be applied to the proposed alternative actions of this relief request. BWR licensees examine reactor internals in accordance with BWRVIP guidelines. These guidelines are written for the safety significant vessel internal components and provide appropriate examination and evaluation criteria using appropriate methods and re-examination frequencies. The BWRVIP has established a reporting protocol for examination results and deviations. The NRC has agreed with the BWRVIP approach as documented in References 1through15. As additional justification, Enclosure 1, "Comparison of ASME Section XI Examination Requirements to BWRVIP Examination Requirements," provides specific examples that compare the inspection requirements of ASME Section XI Item Numbers Bl3.10, B13.20, B13.30, and B13.40 in Table IWB-2500-1, to the inspection requirements in the BWRVIP documents. Specific BWRVIP documents are provided as examples. This comparison also includes a discussion of the inspection methods. These comparisons demonstrate that use of these guidelines, as an alternative to the subject ASME Section XI requirements, provide an acceptable level of quality and safety and will not adversely impact the health and safety of the public. The BWRVIP provides BWR Vessel and Internals Inspection Summaries to the NRC periodically. Table 1 contains the BWR Vessel and Internals Inspection Summaries transmitted to the NRC that includes Peach Bottom Atomic Power Station, Units 2 and 3. These summaries provide, on a component-by-component basis, the examination methods utilized, the examination frequency to date, and the results of the examinations during the previous interval. These summaries also contain the identified corrective actions. This information reflects the compilation of the BWRVIP outage reports.

10 CFR 50.SSa Relief Request ISR-03 Revision 0 (Page 4 of 17) Corrective actions and examinations performed prior to the BWRVIP were implemented to the requirements of ASME Section XI, as applicable. Table 1 BWR Vessel and Internals Inspection Summaries Unit Accession Number Document Title Document Date Project No. 704-BWR Peach Bottom Vessel and Internals Atomic Power ML17187A190 Inspection Summaries for 06/30/2017 Station, Unit 2 Fall 2016 Outages (Reference 19) Project No. 704-BWR Peach Bottom Vessel and Internals Atomic Power ML16189A217 Inspection Summaries for 06/29/2016 Station, Unit 3* Fall 2015 Outages (Reference 20)

• Note the BWR Vessel and Internals Inspection Summary that includes the latest Peach Bottom Atomic Power Station, Unit 3 outage in Fall 2017 has not been assembled and transmitted to the NRC by the BWRVIP.

When a BWRVIP guideline refers to ASME Section XI, the technical requirements of ASME Section XI as described by the BWRVIP guideline will be met, but the examination is under the auspices of the BWRVIP program as defined by BWRVIP-94NP-R2, "BWR Vessel and Internals Project, Program Implementation Guide." The reactor vessel internals inspection program at Peach Bottom Atomic Power Station, Units 2 and 3 has been developed and implemented to satisfy the requirements of BWRVIP-94NP-R2. It is recognized that the BWRVIP executive committee periodically revises the BWRVIP guidelines to address industry operating experience, include enhancements to inspection techniques, and add or adjust flaw evaluation methodologies. BWRVIP-94NP-R2 states that where guidance in existing BWRVIP documents has been supplemented or revised by subsequent correspondence approved by the BWRVIP Executive Committee, the vessel and internals program shall be modified to reflect the new requirements and implement the guidance within two refueling outages, unless a different schedule is specified by the BWRVIP. However, if new guidance approved by the Executive Committee includes changes to NRC approved BWRVIP guidance that are less conservative than those approved by the NRC, the less conservative guidance shall be implemented only after the NRC approves the changes, which generally means publication of a "-A" document or equivalent. Where the revised version of a BWRVIP inspection guideline continues to also meet the

10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 5 of 17) requirements of the version of the BWRVIP inspection guideline that forms the safety basis for the NRC-authorized proposed alternative to the requirements of 10CFR50.55a, it may be implemented. Otherwise, the revised guidelines will only be implemented after NRC approval of the revised BWRVIP guidelines or a plant-specific request for relief has been approved. Peach Bottom Atomic Power Station, Units 2 and 3 are a BWR/4 design. Table 2 compares present ASME Section XI Examination Category B-N-1 and B-N-2 requirements with the above current BWRVIP guideline requirements, as applicable, to Peach Bottom Atomic Power Station, Units 2 and 3. Therefore, Table 2 only represents the most current comparison. Any deviations from the referenced BWRVIP guidelines for the duration of the proposed alternative will be appropriately documented and communicated to the NRC, per the BWRVIP Deviation Disposition Process. Peach Bottom Atomic Power Station currently has one open deviation disposition applicable to this relief request; Table 3 provides a summary of this item. Note that other regulatory commitments (e.g., NUREG-0619, IGSCC) are still implemented separately from the ISi Program or this relief request as Augmented Examination Programs. In the event that conditions are identified that require repair or replacement and the component is within the jurisdiction of ASME Section XI (welded attachments to the reactor vessel or welded core support structure), the repair or replacement activities will be performed in accordance with ASME Section XI, Article IW A-4000. Subsequent examinations will be in accordance with the applicable BWRVIP guideline. As part of the BWRVIP initiative, the BWR reactor internals and attachments were subjected to a safety assessment to identify those components that provide a safety function and to determine if long-term actions were necessary to ensure continued safe operation. The safety functions considered are those associated with (1) maintaining a coolable geometry, (2) maintaining control rod insertion times, (3) maintaining reactivity control, (4) assuring core cooling, and (5) assuring instrumentation availability. The results of the safety assessment are documented in BWRVIP-06-Rl-A, "BWR Vessel and Internals Project, Safety Assessment of BWR Internals," which has been approved by the NRC. As a result of BWRVIP-06-Rl-A, component specific BWRVIP guidelines were developed providing appropriate examination and evaluation requirements to address the specific component safety function and potential degradation mechanism.

6.

Duration of Proposed Alternative Relief is requested for the fifth ISi interval for PBAPS, Units 2 and 3.

7.

Precedents 10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 6 of 17)

1)

The Exelon Generation Company/ AmerGen fleet-wide Relief Request for BWRVIP was authorized conditionally by NRC Safety Evaluation (SE) dated April 30, 2008 (ADAMS Accession No. ML080980311); Relief Request 15R-03 for the PBAPS, Units 2 and 3 fifth ISi interval utilizes a similar approach to the previously approved relief request (Reference 16).

2)

Relief Request 14R-02 was authorized conditionally for LaSalle County Generating Station, Units 1 and 2, by NRC SE dated November 17, 2017 (Reference 17).

3)

Relief Request IR-056, Revision 2 was authorized conditionally for Perry Nuclear Power Plant, Unit 1 by NRC SE dated January 29, 2018 (Reference 18).

8.

References

1)

"TR-105969-R19 (BWRVIP-03) Revision 19: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines," EPRI Technical Report 3002008095, dated December 2016.

2)

Letter from K. Hsueh (NRC) to BWRVIP, "U.S. Nuclear Regulatory Commission Approval Letter for Electric Power Research Institute Topical Report, BWRVIP-18, Revision 2-A, BWR [Boiling Water Reactor] Vessel and Internals Project, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines (TAC No. MF8415)," dated December 21, 2016.

3)

"BWRVIP-06, Revision 1-A: BWR Vessel and Internals Project, Safety Assessment of BWR Reactor Internals," EPRI Technical Report 1019058, dated March 2010.

4)

Letter from NRC to BWRVIP, "Final Safety Evaluation of BWRVIP Vessel and Internals Project, 'BWR Vessel and Internals Project, BWR Core Plate Inspection and Flaw Evaluation Guideline (BWRVIP-25),' EPRI Report TR-107284, December 1996 (TAC No. M97802)," dated December 19, 1999.

5)

Letter NRC to BWRVIP, "NRC Approval Letter of BWRVIP-26-A, BWR Vessel and Internals Project, Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines," dated September 9, 2005.

6)

Letter from NRC to BWRVIP, "Final Safety Evaluation of the 'BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38),' EPRI Report TR-108823 (TAC No. M99638)", dated July 24, 2000.

10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 7 of 17)

7)

"BWRVIP-41, Revision 3: BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 1021000, dated September 2010.

8)

Letter from NRC to BWRVIP, "NRC Approval Letter ofBWRVIP-47-A, 'BWR Vessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines,'" dated September 9, 2005.

9)

Letter from NRC to BWRVIP, "NRC Approval Letter of BWRVIP-48-A, 'BWR Vessel and Internals Project Vessel ID Attachment Weld Inspection and Flaw Evaluation Guideline,"' dated July 25, 2005.

10)

"BWRVIP-49-A: BWR Vessel and Internals Project, Instrument Penetration Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 1006602, dated April 04, 2002.

11)

"BWRVIP-76, Revision 1-A, BWR Vessel and Internals Project, BWR Core Shroud Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002005566, dated April 2015.

12)

Letter from Chairman, BWR Vessel and Internals Project to NRC, "Project No. 704 - BWRVIP Program Implementation Guide (BWRVIP-94NP, Revision 2)," dated September 22, 2011.

13)

"BWRVIP-138, Revision 1-A: BWR Vessel and Internals Project, Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 1025136, dated October 2012.

14)

"BWRVIP-180: BWR Vessel and Internals Project, Access Hole Cover Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 1013402, dated November 2007.

15)

"BWRVIP-183-A: BWR Vessel and Internals Project, Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002010551, dated November 2017.

16)

Letter from R. Gibbs (NRC) to C. G. Pardee (Exelon Generation Company/AmerGen), "Clinton Power Station Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Generating Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1and2 - Relief Request to Use Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (TAC Nos. MD5352 through MD5363)," dated April 30, 2008 (ADAMS Accession No. ML08098031 l).

10 CFR 50.SSa Relief Request ISR-03 Revision 0 (Page 8 of 17)

17)

Letter from D. J. Wrona (NRC) to B. C. Hanson (Exelon), "LaSalle County Station, Units 1 and 2, Relief from the Requirements of the ASME Code and OM Code Re: Relief Requests 14R-02, 14R-03, 14R-06, 14R-07, and 14R-09, Proposed Alternatives to Various Inservice Inspection Interval (ISi) Requirements of the American Society of Mechanical Engineers (ASME Code), Section XI, 2007 Edition with the 2008 Addenda for the Fourth 10-Year ISi Interval (EPID Nos. L-2017-LLR-0038 (CAC Nos. MF9760 and MF9761), L-LR-2017-0076 (CAC Nos. MF9762 and MF9763), L-2017-LLR-0033 (CAC Nos. MF9766 and MF9767), L-2017-LLR-0035 (CAC Nos. MF9770 and MF9771), and L-2017-LLR-0037 (CAC Nos. MF9768, and MF9769))," Dated November 17, 2017 (ADAMS Accession No. MLl 7305B279).

18)

Letter from D. J. Wrona (NRC) to D. B. Hamilton (FirstEnergy Nuclear Operating Company), "Perry Nuclear Power Plant, Unit No. 1 - Approval of Alternative to Use BWRVIP Guidelines in Lieu of Certain ASME Code Requirements (CAC No. MG0149; EPID 2017-LLR-0112) (L-17-183)," dated January 29, 2018 (ADAMS Accession No. ML18023A625).

19)

Letter 2017-081 from BWRVIP to NRC, "Project No. 704 - BWR Vessel and Internals Inspection Summaries for Fall 2016 Outages," dated June 30, 2017 (ADAMS Accession No. ML17187Al90).

20)

Letter 2016-071 from BWRVIP to NRC, "Project No. 704 - BWR Vessel and Internals Inspection Summaries for Fall 2015 Outages," dated June 29, 2016 (ADAMS Accession No. ML16189A217).

ASME Section XI Item Number, Table IWB-2500-1 B13.10 Reactor Vessel Interior B13.20 Interior Attachments Within Beltline Region Bl3.30 Interior Attachments Beyond Beltline 10 CFR 50.SSa Relief Request ISR-03 Revision 0 (Page 9 of 17) TABLE2 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Exam Requirements with BWRVIP Guidance Requirements1 ASME ASME ASME BWRVIP Authorized BWRVIP BWRVIP Component Section XI Section XI Section XI Alternative Exam Exam Frequency Scope Exam Frequency Scope Reactor Vessel Accessible VT-3 Each BWRVIP Overview examinations of components during Interior Areas period R2-A, 25, 26-A, BWRVIP examinations are preformed to satisfy 38, 41-R3, 47-A, ASME Section XI VT-3 visual examination 48-A, 76-Rl-A, requirements. 138-Rl-A, 180, and 183-A Jet Pump Riser Accessible VT-1 Each BWRVIP-48-A, Riser Brace EVT-1 25% each 6 years. Braces Welds 10-year Table 3-2 Attachment Lower Interval BWRVIP-48-A, Bracket VT-1 Each 10-year Surveillance Table 3-2 Attachment Interval. Specimen Holder Brackets Steam Dryer Hold-Accessible VT-3 Each BWRVIP-48-A, Bracket VT-3 Each 10-year down Brackets Welds 10-year Table 3-2 Attachment Interval. Guide Rod Interval BWRVIP-48-A, Bracket VT-3 Each 10-year Brackets Table 3-2 Attachment Interval. Steam Dryer BWRVIP-48-A, Bracket EVT-1 Each 10-year Support Brackets Table 3-2 Attachment Interval. Feedwater Sparger BWRVIP-48-A, Bracket EVT-1 Each 10-year Brackets Table 3-2 Attachment Interval. Core Spray Piping BWRVIP-48-A, Bracket EVT-1 100% every 4 Brackets Table 3-2 Attachment refueling cycles. Upper BWRVIP-48-A, Bracket VT-3 Each 10-year Surveillance Table 3-2 Attachment Interval. Specimen Holder Brackets

Shroud Support (Weld H9) Shroud Support Legs (Weld Hl2) (Rarely Accessible) B 13.40 Shroud Support Accessible Core Support (Welds H8 and Surfaces Structure H9) Shroud Support Legs (Weld Hl2) (Rarely Accessible) Shroud Horizontal Welds Shroud Vertical Welds 10 CFR 50.SSa Relief Request ISR-03 Revision 0 (Page 10 of 17) BWRVIP-38, 3.3, Figure 3-5 BWRVIP-38, 3.2.3 VT-3 Each BWRVIP-38, 10-year 3.3, Interval Figures 3-4 and 3-5 BWRVIP-38, 3.2.3 BWRVIP Rl-A, 2.2, Figure 2-3 BWRVIP Rl-A, 2.3, 3.3, Figures 2-4, 2-5, and 3-2 Weld H92 EVT-1 or UT Based on as-found conditions, to a maximum of6 years for EVT-1, 10 years for UT. Weld Hl2 Per When accessible. BWRVIP-38 NRCSE (7 /24/2000), inspect with appropriate method3 Welds H8 EVT-1 or UT Based on as found and H92 conditions, to a maximum of6 years for EVT-1, 10 years for UT. Weld Hl2 Per When accessible BWRVIP-38 NRCSE (7 /24/2000), inspect with appropriate method3 Welds Hl-EVT-1 or UT Based on as found H7 as conditions, to a applicable maximumof6 years for one-sided EVT-1, 10 years for UT. Vertical EVT-1 or UT Maximum of6 Welds as years for one-sided applicable EVT-1, 10 years for UT.

NOTES: 10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 11 of 17)

1) This table provides only an overview of the requirements. For more details, refer to ASME Section XI, Table IWB-2500-1, and the appropriate BWRVIP document.

2)

In accordance with Appendix A of BWRVIP-38, a site specific evaluation will determine the minimum required weld length to be examined.

3) When inspection tooling and methodologies are available, they will be utilized to establish a baseline inspection of these welds.

PLANT BWRVIP DOCUMENT Peach Bottom BWRVIP-25 Atomic Power Station, Units 2 and 3 10 CFR 50.SSa Relief Request ISR-03 Revision 0 (Page 12 of 17) TABLE3 BWRVIP Deviations LETTER DATE TO NRC DEVIATION Letter from S. Kuczynski BWRVIP-25 requires ultrasonic or visual examination of (Exelon Generation Company, the core plate bolts. There are currently no examination LLC) to U.S. Nuclear methods available for these bolts and is being addressed as Regulatory Commission, a BWRVIP generic issue. Analytical evaluation has been "Deviation from BWR Vessel completed to support operation. and Internals Project (BWRVIP) Guiddine Inspection of the Core Plate Bolts." dated March 31, 2011 (RS-11-053) (Accession No. MLI 10910333) APPLICABILITY This deviation does not impact the basis for the use of this relief request.

10 CFR 50.SSa Relief Request ISR-03 Revision 0 (Page 13 of 17) Comparison of Code Examination Requirements to BWRVIP Examination Requirements The following paragraphs provide a comparison of the examination requirements in ASME Section XI, Table IWB-2500-1, Item Numbers B13.10, B13.20, B13.30, and B13.40, to the examination requirements in the BWRVIP guidelines. Specific BWRVIP guidelines are provided as examples for comparisons. This comparison also includes a discussion of the examination methods.

1.

ASME Section XI Requirement* B13.10 - Reactor Vessel Interior Accessible Areas (B-N-1) ASME Section XI requires a VT-3 visual examination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during normal refueling outages. The frequency of these examinations is specified as the first refueling outage, and at intervals of approximately three years, during the first ISi interval, and each inspection period during each successive 10-year ISi Interval. Typically, these examinations are performed every inspection period during the ten-year ISi Interval. This examination requirement is a non-specific requirement that is a departure from the traditional ASME Section XI examinations of welds and surfaces. As such, this requirement has been interpreted and satisfied differently across the domestic fleet. The purpose of the examination is to identify relevant conditions such as distortion or displacement of parts; loose, missing, or fractured fasteners; foreign material, corrosion, erosion, or accumulation of corrosion products, wear, and structural degradation. Portions of the various examinations required by the applicable BWRVIP guidelines require examination of accessible areas of the reactor vessel during refueling outages. Examination of core spray piping and spargers (BWRVIP-18-R2-A), core plate (BWRVIP-25), top guide (BWRVIP-26-A), jet pump welds and components (BWRVIP-41-R3), interior attachments (BWRVIP-48-A), core shroud welds (BWRVIP-76-Rl-A), shroud support (BWRVIP-38), access hole cover (BWRVIP-180), top guide grind beams (BWRVIP-183-A), and lower plenum components (BWRVIP A) provides such access. Locating and examining specific welds and components within the reactor vessel areas above, below (if accessible), and surrounding the core (annulus area) entails access by remote camera systems that essentially perform equivalent VT-3 visual examination of these areas or spaces as the specific weld or component examinations are performed. This provides an equivalent method of visual examination on a more frequent basis than that required by ASME Section XI. Evidence of wear, structural degradation, loose, missing, or displaced parts, foreign materials, and corrosion product buildup can be, and has been observed during the course of implementing these BWRVIP examination requirements. Therefore, the specified BWRVIP guideline requirements meet or exceed the subject ASME Section XI requirements (including method and frequency requirements) for examination of the interior of the reactor vessel. Accordingly, these BWRVIP

10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 14 of 17) examination requirements provide an acceptable level of quality and safety as compared to the subject ASME Section XI requirements.

2.

ASME Section XI Requirement - B13.20 - Interior Attachments Within the Beltline CB-N-2) ASME Section XI requires a VT-1 visual examination of accessible reactor vessel interior surface attachment welds within the beltline each 10-year interval. In the General Electric - Hitachi BWR/4 design, this includes the jet pump riser brace welds-to-reactor vessel wall and the lower surveillance specimen support bracket welds-to-reactor vessel wall. In comparison, the BWRVIP requires the same examination method and frequency for the lower surveillance specimen support bracket welds, and requires an enhanced VT-1 (EVT-1) visual examination of the remaining attachment welds in the beltline region in the first 12 years, and then 25% during each subsequent 6 years. The jet pump riser brace examination requirements are provided below to show a comparison between ASME Section XI and the BWRVIP examination requirements. Comparison to BWRVIP Requirements - Jet Pump Riser Braces CBWRVIP-41-R3 and BWRVIP-48-A) ASME Section XI requires a 100% VT-1 visual examination of the jet pump riser brace-to-reactor vessel wall pad welds each 10-year interval. The BWRVIP-41-R3 requires an EVT-1 visual examination of the jet pump riser brace-to-reactor vessel wall pad welds in the first 12 years, and then 25% during each subsequent 6 years. BWRVIP-48-A specifically defines the susceptible regions of the attachment that are to be examined. ASME Section XI VT-1 visual examination is conducted to detect discontinuities and imperfections on the surfaces of components, including such conditions as cracks, wear, corrosion, or erosion. The BWRVIP EVT-1 visual examination is conducted to detect discontinuities and imperfections on the surface of components and is additionally specified to detect potentially very tight cracks characteristic of fatigue and intergranular stress corrosion cracking (IGSCC), the relevant degradation mechanisms for these components. General wear, corrosion, or erosion although generally not a concern for inherently tough, corrosion resistant stainless steel material, would also be detected during the process of performing a BWRVIP EVT-1 visual examination. The ASME Section XI VT-1 visual examination method requires that a letter character with a maximum height of 0.044 inches be read at a maximum distance of 2 feet. The BWRVIP EVT-1 visual examination method requires resolution of 0.044 inch characters on the examination surface and additionally the performance of a cleaning assessment

10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 15 of 17) and cleaning as necessary. BWRVIP-48-A includes a diagram for the configuration and prescribes examination for each configuration including Peach Bottom Atomic Power Station, Units 2 and 3. The calibration standards used for BWRVIP EVT-1 visual examinations utilize the ASME Section XI characters, thus assuring at least equivalent resolution compared to the ASME Section XI requirements. Although the BWRVIP examination may be less frequent, it is a more comprehensive method. Therefore, the enhanced flaw detection capability of an EVT-1 visual examination with a less frequent examination schedule provides an acceptable level of quality and safety to that provided by ASME Section XI.

3.

ASME Section XI Requirement

• B13.30 - Interior Attachment Beyond the Beltline Region (B-N-2)

ASME Section XI requires a VT-3 visual examination of accessible reactor vessel interior surface attachment welds beyond the beltline each 10-year interval. In the BWR/4 design, this includes the core spray piping primary and supplemental support bracket welds-to-reactor vessel wall, the upper surveillance specimen support bracket welds-to-reactor vessel wall, the feedwater sparger support bracket welds-to-reactor vessel wall, the steam dryer support and hold down bracket welds-to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, the shroud support plate-to-reactor vessel wall, and the shroud support gussets. BWRVIP-48-A requires as a minimum the same VT-3 visual examination method as ASME Section XI for some of the interior attachment welds beyond the beltline region, and in some cases specifies an EVT-1 for these welds. For those interior attachment welds that have the same VT-3 method of visual examination, the same scope of examination (accessible welds), the same examination frequency (each 10-year interval), and the same ASME Section XI flaw evaluation criteria are used. Therefore, the level of quality and safety provided by the BWRVIP requirements are equivalent to that provided by ASME Section XL For the core spray piping attachment welds, the steam dryer support bracket attachment welds, the feedwater sparger support bracket attachment welds, and the shroud support plate-to-reactor vessel welds, the BWRVIP guidelines require an EVT-1 visual examination at the same frequency as ASME Section XI, or at a more frequent rate. Therefore, the BWRVIP enhanced examination requirements provide the same level of quality and safety compared to that provided by ASME Section XI. The core spray piping bracket-to-reactor vessel attachment weld is used as an example for comparison between ASME Section XI and BWRVIP examination requirements as discussed below. Comparison to BWRVIP Requirements - Core Spray Piping Bracket Welds CBWRVIP-48-A) The ASME Section XI examination requirement is a VT-3 visual examination of each weld every 10 years.

10 CFR 50.SSa Relief Request ISR-03 Revision 0 (Page 16 of 17) The BWRVIP-48-A visual examination requirement is an EVT-1 for the core spray piping bracket attachment welds with each weld examined every four cycles (8 years for units with a two year fuel cycle). The BWRVIP-48-A visual examination method EVT-1 has superior flaw detection and sizing capability, the examination frequency is the same or greater than ASME Section XI requirements, and the same flaw evaluation criteria are used. ASME Section XI VT-3 visual examination is conducted to detect component structural integrity by ensuring the components general condition is acceptable. An EVT-1 visual examination is conducted to detect discontinuities and imperfections on the examination surfaces, including such conditions as tight cracks caused by IGSCC or fatigue, the relevant degradation mechanisms for BWR internal attachments. Therefore, because the EVT-1 visual examination method provides the same examination scope (accessible welds), a similar or increased examination frequency in most cases, and the same flaw evaluation criteria as ASME Section XI, and the level of quality and safety provided by the BWRVIP criteria exceeds that provided by the ASME Section XI requirements.

4.

ASME Section XI Requirement - B13.40 - Core Support Structures {B-N-2) ASME Section XI requires a VT-3 visual examination of accessible surfaces of the reactor vessel core support structure each 10-year interval. In the BWR/4 design, the core support structure has primarily been considered the shroud support structure, including the shroud support plate (annulus floor), the shroud support ring, the shroud support welds, and the shroud support legs (if accessible). Historically, this requirement has been interpreted and satisfied differently across the industry. The proposed alternate examinations replace this ASME Section XI requirement with specific BWRVIP guidelines that examine susceptible locations for known relevant degradation mechanisms. Comparison to BWRVIP Requirements - Shroud Supports (BWRVIP-38) ASME Section XI requires a VT-3 visual examination of accessible surfaces each 10-year interval. The BWRVIP-38 requires, an EVT-1every6 years or ultrasonic examination (UT) every 10 years. BWRVIP recommended examinations of core support structures are focused on the known susceptible areas of this structure, including the welds and associated weld heat affected zones. In many locations, the BWRVIP guidelines require a volumetric examination of the susceptible welds at the same frequency identical to the ASME Section XI requirement.

10 CFR 50.55a Relief Request ISR-03 Revision 0 (Page 17 of 17) The BWRVIP guidelines require an EVT-1 or UT of core support structures. The core shroud is used as an example for comparison between the ASME Section XI and BWRVIP examination requirements as shown below. Comparison to BWRVIP Requirements - Core Shroud CBWRVIP-76-Rl-A) ASME Section XI requires a VT-3 visual examination of accessible surfaces every 10 years. The BWRVIP-76-Rl-A requires an EVT-1 visual examination from the inside and outside surface, where accessible, or UT of select circumferential welds that have not been structurally replaced with a shroud repair, at a calculated "end of interval" that will vary depending upon the amount of flaws present, but not to exceed 10 years. Therefore, the BWRVIP referenced examinations are the same or superior to ASME Section XI requirements. Shroud vertical welds are recommended in BWRVIP-76-Rl-A and have the same basic VT-3 method of visual examination or better, the same examination frequency (each 10-year interval) and comparable flaw evaluation criteria. Therefore, the BWRVIP requirements provide a level of quality and safety equivalent to that provided by ASME Section XI. For other core support structure components, the BWRVIP requires an EVT-1 visual examination or volumetric examination (UT) of core support structures. Summary The BWRVIP recommended examinations specify locations that are known to be vulnerable to BWR relevant degradation mechanisms rather than accessible surfaces. The BWRVIP examination methods (EVT-1 or UT) are superior to the ASME Section XI required VT-3 visual examination for flaw detection and characterization. In most cases, the BWRVIP examination frequency is equivalent to or more frequent than the examination frequency required by the ASME Section XI. In cases where the BWRVIP examination frequency is less frequent than required by ASME Section XI, the BWRVIP examinations are performed in a more comprehensive manner and focus on the areas most vulnerable. Therefore, the superior flaw detection and characterization capability, with an equivalent or more frequent examination frequency, or with a less frequent examination frequency but with those examinations being performed in a more comprehensive manner, and using comparable flaw evaluation criteria, results in the BWRVIP criteria providing a level of quality and safety equivalent to or superior to that required by the ASME Section XI requirements.

10 CFR 50.55a Relief Request ISR-04 Revision 0 (Page 1 of 11) Request for Relief ISR-04 to Implement an Alternative Concerning Nozzle-to-Vessel Weld and Inner Radii Examination Requirements in Accordance with 10 CFR 50.SSa(z)(l)

1.

ASME Code Component(s) Affected Code Class:

Reference:

Examination Category: Item Number:

== Description:== Component Numbers: 1 IWB-2500, Table IWB-2500-1 B-D B3.90 and B3.100 Alternative to IWB-2500, Table IWB-2500-1 for Nozzle-to-Vessel Weld and Inner Radii Examination Requirements in Accordance with ASME Code Case N-702 Reactor Vessel Nozzles: N2, N3, NS, N6, NS (See Enclosures 1 and 2 for complete list of nozzle identifications)

2.

Applicable Code Edition and Addenda

The fifth ten-year interval of the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Inservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2013 Edition. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2013 Edition is implemented.

3.

Applicable Code Requirement

The applicable requirement is contained in Table IWB-2500-1, "Examination Category B-D, Full Penetration Welded Nozzles in Vessels." Class 1 Reactor Vessel nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Number B3.90, "Nozzle-to-Vessel Welds," and B3.100, "Nozzle Inside Radius Section." The required method of examination is volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles are examined each interval. All of the nozzle assemblies identified in Enclosures 1 and 2 are full penetration welds.

4.

Reason for Request

NRC Regulatory Guide 1.147, Revision 18 conditionally accepts the use of ASME Code Case N-702. This code case provides an alternative to performing examination of 100% of the nozzle-to-vessel welds and inner radii for Examination Category B-D nozzles with the exception of the Feed water and Control Rod Drive Return Line (CRDRL) nozzles.

10 CFR 50.55a Relief Request ISR-04 Revision 0 (Page 2 of 11) The alternative is to perform examination of a minimum of 25% of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size, excluding the Feedwater and CRDRL Nozzles. Code Case N-702 has been approved for use in Regulatory Guide 1.147, Revision 18 with conditions as noted below: The technical basis supporting the implementation of this Code Case is addressed by BWRVIP-108: BWR Vessel and Internals Project, "Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1003557, October 2002 (ML023330203) and BWRVIP-241: BWR Vessel and Internals Project, "Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1021005, October 2010 (ML11119A041). The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated December 18, 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013(ML13071A240) are met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRC prior to the application of the Code Case. The analyses in BWRVIP-108NP and BWRVIP-241 were based on predicted fatigue crack growth over the initial licensed operating period and assumed additional fatigue cycles in evaluating fatigue crack growth. Previous BWRVIP documents have demonstrated that stress corrosion crack (SCC) growth represents the majority of the crack growth and that crack growth due to additional mechanical/thermal fatigue cycles introduced by the extended operation time is insignificant compared to hypothetical sec growth. Thus, the amount of thermal cycle driven fatigue crack growth due to extended operation to 60 years is not a controlling factor in the probability of failure of the BWR reactor vessel nozzles. The proposed alternative provides an acceptable level of quality and safety based on the technical content of BWRVJP-108 and BWRVIP-241, as endorsed by the NRC Safety Evaluations.

5.

Proposed Alternative and Basis for Use In accordance with 10CFR50.55a(z)(l), relief is requested from performing the required examinations on 100 percent of the nozzle assemblies identified in Tables 5-1 and 5-2 below (see Enclosures 1 and 2 for a list of RPV Examination Category B-D nozzles applicable to this relief request). As an alternative, for all welds and inner radii identified in Tables 5-1and5-2, PBAPS, Units 2 and 3 proposes to examine a minimum of 25 percent of the nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with ASME Code Case

10 CFR 50.55a Relief Request ISR-04 Revision 0 (Page 3 of 11) N-702. For the nozzle assemblies identified in Enclosures l and 2, this would mean 25 percent from each of the groups identified in Tables 5-1 and 5-2 during the 120-month interval. Table 5-1 Peach Bottom Atomic Power Station, Unit 2 RPV Examination Category B-D Nozzle Summary Total Minimum Comments Group Number Number to be Results1 Examined Recirculation Inlet 10 3 Two (2) nozzles were inspected (N2) in the fourth ISi interval. No recordable indications. One (1) is currently scheduled to be inspected within the fourth ISi interval. Main Stearn 4 1 One ( 1) nozzle is scheduled to (N3) be inspected in the fourth ISi interval. Core Spray 2 1 One ( 1) nozzle was inspected in (NS) the fourth ISi interval. No recordable indications. Nozzles On Top Head 2 1 One ( 1) nozzle was inspected in (N6) the fourth ISi interval. No recordable indications. Jet Pump Instrument 2 1 One ( 1) nozzle was inspected in (NS) the fourth ISi interval. No recordable indications. Note:

1. The nozzle-to-vessel weld and inner radius examinations are performed together.

10 CFR 50.55a Relief Request 15R-04 Revision 0 (Page 4 of 11) Table 5-2 Peach Bottom Atomic Power Station, Unit 3 RPV Examination Category B-D Nozzle Summary Total Minimum Comments Group Number Number to be Results1 Examined Recirculation Inlet 10 3 Three (3) nozzles were inspected (N2) in the fourth ISI interval. No recordable indications. Main Steam 4 1 One ( 1) nozzle was inspected in (N3) the fourth ISI interval. No recordable indications. Core Spray 2 1 One (1) nozzle was inspected in (NS) the fourth ISI interval. No recordable indications. Nozzles On Top Head 2 1 One ( 1) nozzle was inspected in (N6) the fourth ISI interval. No recordable indications. Jet Pump Instrument 2 1 One (1) nozzle was inspected in (N8) the fourth ISI interval. No recordable indications. Note:

1. The nozzle-to-vessel weld and inner radius examinations are performed together.

The examinations in Tables 5-1 and 5-2 will be scheduled in accordance with ASME Section XI, IWB-2411, "Inspection Program." ASME Code Case N-702 stipulates that a VT-1 visual examination may be used in lieu of the volumetric examination for the inner radii (i.e., Item Number B3.100, "Nozzle Inside Radius Section"). This VT-1 examination is outlined in Code Case N-648-1 ("Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles Section XI, Division l "). EGC will perform either volumetric examination or VT-1 examination of the inner radius as required by ASME Code Case N-702.

10 CFR 50.SSa Relief Request ISR-04 Revision 0 (Page 5 of 11) Electric Power Research Institute (EPRI) Technical Report (TR) 1003557, "BWRVJP-108: Boiling Water Reactor Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a Low Temperature Overpressure (LTOP) event are very low (i.e., <1 x 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25 percent of each nozzle type is technically justified. EPRI Report BWRVJP-241 received a final NRC Safety Evaluation Report on April 19, 2013 (ML13071A240). In the NRC Safety Evaluation Report, Section 5.0, "Conditions and Limitations," indicates that each licensee who plans to request relief from ASME Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radii sections may reference the BWRVJP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVJP-241 report to their units in the relief request by demonstrating that the following general and nozzle-specific criteria are satisfied: In the case of Peach Bottom Atomic Power Station, Units 2 and 3, the single set of values (e.g., nozzle radii, nozzle thicknesses, etc.) used in the following equations are correct and applicable to Peach Bottom Atomic Power Station, Units 2 and 3. These values are minimum design values.

1.

The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour. PBAPS, Units 2 and 3 Technical Specification (TS) Surveillance Requirement (SR) 3.4.9.1, "RCS Pressure and Temperature (PIT) Limits," provides a Reactor Coolant System (RCS) heatup/cooldown rate as specified in the Pressure Temperature Limits Report. This report limits the RCS heatup/cooldown rate to less than or equal to 100°F in any 1-hour period. The heatup/cooldown rate is also referenced in the Peach Bottom Atomic Power Station operating procedures.

2.

For the Recirculation Inlet Nozzles (N2), the following criteria must be met:

a. (pr/t)ICRPV :$ 1.15; p=RPV Normal Operating Pressure r=RPV inner radius t=RPV wall thickness CRPV=

(pr/t)/CRPv = 1.097 :$ 1.15 1035 psi 125.5 in. 6.125 in. 19332 The calculation for the Peach Bottom Atomic Power Station, Units 2 and 3, N2 Nozzle results in a maximum value of 1.097, which satisfies this criteria.

10 CFR 50.55a Relief Request ISR-04 Revision 0 (Page 6 of 11) p=RPV Normal Operating Pressure ro=nozzle outer radius ri=nozzle inner radius CNOZZLE 1035 psi 12.5 in. 5.784 in. 1637 The calculation for the Peach Bottom Atomic Power Station, Units 2 and 3, N2 Nozzle results in a maximum value of 0.977, which satisfies this criteria.

3.

For the Recirculation Outlet Nozzles (Nl), the following criteria must be met:

a. (pr/t)/CRPv:=: 1.15; p=RPV Normal Operating Pressure r=RPV inner radius t=RPV wall thickness CRPV=

(pr/t)/CRPv = 1.311>1.15 1035 psi 125.5 in. 6.125 in. 16171 The calculation for the Peach Bottom Atomic Power Station, Units 2 and 3 Nl Nozzle results in a value of 1.311, which is higher than 1.15 and does not meet the criteria; therefore, the N 1 nozzles are not included in this relief request. p=RPV Normal Operating Pressure ro=nozzle outer radius fi=nozzle inner radius CNOZZLE 1035 psi 26.5 in. 12.97 in. 1977 The calculation for the Peach Bottom Atomic Power Station, Units 2 and 3 Nl Nozzle results in 0.853, which is less than 1.59. Based upon the above information, all PBAPS RPV nozzle-to-vessel shell full penetration welds and nozzle inner radii sections, with the exception of the Recirculation Outlet Nozzles, meet the general and nozzle-specific criteria in BWRVIP-241. The Recirculation Outlet (Nl) Nozzles are not included in this relief request, and both nozzles will be examined during the fifth ISi interval.

10 CFR 50.55a Relief Request ISR-04 Revision 0 (Page 7 of 11) The analyses for the nozzles in BWRVIP-108NP and BWRVIP-241 are based on the assumption that fluence at the nozzles is negligible because the analysis is for the initial 40 years of plant operation and do not address an extended operating period. Pressure-Temperature Limits reports applicable to Peach Bottom Atomic Power Station, Units 2 and 3, concluded that peak fluence over the period of extended operation (54 effective full power years) is expected to be less than the fluence criteria of l.OE17 n/cm2, as contained in 1 OCFR50, Appendix H for all nozzles and welds for which this relief request is applied. Therefore, the fluence criteria is satisfied and use of BWRVIP-108 and BWRVIP-241 remain applicable to the Peach Bottom Atomic Power Station nozzles contained in this relief request. Therefore, the use of ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 10CFR50.55a(z)(l) for all applicable full penetration RPV nozzle-to-vessel shell welds and nozzle inner radii sections for the Fifth ISi Interval, with the exception of the Recirculation Outlet Nozzles.

6.

Duration of Proposed Alternative Relief is requested for the fifth ISi interval for PBAPS, Units 2 and 3, or until the NRC approves Code Case N-702, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

7.

Precedents PBAPS, Units 2 and 3, fourth ISi interval Relief Request 14R-52 was authorized by NRC Safety Evaluation (SE) dated January 24, 2012 (ADAMS Accession No. MLl 12770217). Relief Request 15R-04 for the PBAPS, Units 2 and 3, fifth ten-year ISi interval, utilizes a similar approach to the previously approved relief request.

8.

References

1)

ASME Section XI Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section XI, Division l."

2)

NRC Regulatory Guide 1.14 7, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division l," Revision 18.

10 CFR 50.55a Relief Request ISR-04 Revision 0 (Page 8 of 11) Applicable Peach Bottom Atomic Power Station, Unit 2 Nozzles Component ID Examination Item System Nominal Category Number Pipe Size N2A B-D B3.90 Recirc Inlet 12" N2A-IRS B-D B3.100 Recirc Inlet 12" N2B B-D B3.90 Recirc Inlet 12" N2B-IRS B-D B3.100 Recirc Inlet 12" N2C B-D B3.90 Recirc Inlet 12" N2C-IRS B-D B3.100 Recirc Inlet 12" N2D B-D B3.90 Recirc Inlet 12" N2D-IRS B-D B3.100 Recirc Inlet 12" N2E B-D B3.90 Recirc Inlet 12" N2E-IRS B-D B3.100 Recirc Inlet 12" N2F B-D B3.90 Recirc Inlet 12" N2F-IRS B-D B3.100 Recirc Inlet 12" N2G B-D B3.90 Recirc Inlet 12" N2G-IRS B-D B3.100 Recirc Inlet 12" N2H B-D B3.90 Recirc Inlet 12" N2H-IRS B-D B3.100 Recirc Inlet 12" N2J B-D B3.90 Recirc Inlet 12" N2J-IRS B-D B3.100 Recirc Inlet 12" N2K B-D B3.90 Recirc Inlet 12" N2K-IRS B-D B3.100 Recirc Inlet 12" N3A B-D B3.90 Main Steam 26" N3A-IRS B-D B3.100 Main Steam 26" N3B B-D B3.90 Main Steam 26" N3B-IRS B-D B3.100 Main Steam 26" N3C B-D B3.90 Main Steam 26" N3C-IRS B-D B3.100 Main Steam 26" N3D B-D B3.90 Main Steam 26" N3D-IRS B-D B3.100 Main Steam 26" N5A B-D B3.90 Core Spray 10" N5A-IRS B-D B3.100 Core Spray 10" N5B B-D B3.90 Core Spray 10" N5B-IRS B-D B3.100 Core Spray 10" CH-NA (N6A) B-D B3.90 Head Spray 6" CH-NA-IRS B-D B3.100 Head Spray 6" (N6A) CH-NC (N6B) B-D B3.90 Head Spray 6"

10 CFR 50.SSa Relief Request ISR-04 Revision 0 (Page 9 of 11) Applicable Peach Bottom Atomic Power Station, Unit 2 Nozzles Component ID Examination Item System Nominal Cate2ory Number Pipe Size CH-NC-IRS B-D B3.100 Head Spray 6" (N6B) N8A B-D B3.90 Jet Pump 4" Instrumentation N8A-IRS B-D B3.100 Jet Pump 4" Instrumentation N8B B-D B3.90 Jet Pump 4" Instrumentation N8B-IRS B-D B3.100 Jet Pump 4" Instrumentation

• IRS - Inner Radius Section

10 CFR 50.SSa Relief Request ISR-04 Revision 0 (Page 10 of 11) Applicable Peach Bottom Atomic Power Station, Unit 3 Nozzles Component ID Examination Item System Nominal Cate2ory Number Pipe Size N2A B-D B3.90 Recirc Inlet 12" N2A-IRS B-D B3.100 Recirc Inlet 12" N2B B-D B3.90 Recirc Inlet 12" N2B-IRS B-D B3.100 Recirc Inlet 12" N2C B-D B3.90 Recirc Inlet 12" N2C-IRS B-D B3.100 Recirc Inlet 12" N2D B-D B3.90 Recirc Inlet 12" N2D-IRS B-D B3.100 Recirc Inlet 12" N2E B-D B3.90 Recirc Inlet 12" N2E-IRS B-D B3.100 Recirc Inlet 12" N2F B-D B3.90 Recirc Inlet 12" N2F-IRS B-D B3.100 Recirc Inlet 12" N2G B-D B3.90 Recirc Inlet 12" N2G-IRS B-D B3.100 Recirc Inlet 12" N2H B-D B3.90 Recirc Inlet 12" N2H-IRS B-D B3.100 Recirc Inlet 12" N2J B-D B3.90 Recirc Inlet 12" N2J-IRS B-D B3.100 Recirc Inlet 12" N2K B-D B3.90 Recirc Inlet 12" N2K-IRS B-D B3.100 Recirc Inlet 12" N3A B-D B3.90 Main Steam 26" N3A-IRS B-D B3.100 Main Steam 26" N3B B-D B3.90 Main Steam 26" N3B-IRS B-D B3.100 Main Steam 26" N3C B-D B3.90 Main Steam 26" N3C-IRS B-D B3.100 Main Steam 26" N3D B-D B3.90 Main Steam 26" N3D-IRS B-D B3.100 Main Steam 26" N5A B-D B3.90 Core Spray 10" NS A-IRS B-D B3.100 Core Spray 10" NSB B-D B3.90 Core Spray 10" N5B-IRS B-D B3.100 Core Spray 10" CH-NA (N6A) B-D B3.90 Head Spray 6" CH-NA-IRS B-D B3.100 Head Spray 6" (N6A) CH-NC (N6B) B-D B3.90 Head Spray 6" CH-NC-IRS B-D B3.100 Head Spray 6" (N6B)

10 CFR 50.SSa Relief Request ISR-04 Revision 0 (Page 11 of 11) Applicable Peach Bottom Atomic Power Station, Unit 3 Nozzles Component ID Examination Item System Nominal Category Number Pipe Size N8A B-D B3.90 Jet Pump 4" Instrumentation N8A-IRS B-D B3.100 Jet Pump 4" Instrumentation N8B B-D B3.90 Jet Pump 4" Instrumentation N8B-IRS B-D B3.100 Jet Pump 4" Instrumentation

• IRS - Inner Radius Section

10 CFR 50.55a Relief Request ISR-05 Revision 0 (Page 1 of 11) Request for Relief 15R-05 for the Use of Encoded Phases Array Ultrasonic Examination Techniques in Lieu of Radiography in Accordance with 10 CFR 50.55a(z)(l)

1.

ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel (BPV) Code, Section XI, ISi ferritic piping butt welds requiring radiography during repair/replacement activities.

2.

Applicable Code Edition and Addenda

The fifth 10-year interval of the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Inservice Inspection (ISi) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2013 Edition.

3.

Applicable Code Requirement

10CFR50.55a(b)(2)(xx)(B) requires that "The NDE provision in IW A-4540(a)(2) of the 2002 Addenda of ASME Section XI must be applied when performing system leakage tests after repair and replacement activities performed by welding or brazing on a pressure retaining boundary using the 2003 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)( l)(ii) of this section." IW A-4540(a)(2) of the 2002 Addenda of ASME Section XI requires that the nondestructive examination method and acceptance criteria of the 1992 Edition or later of ASME Section ID be met prior to return to service in order to perform a system leakage test in lieu of a system hydrostatic test. The examination requirements for ASME Section III, circumferential butt welds are contained in ASME Section ill, Subarticles NB-5200, NC-5200, and ND-5200. The acceptance standards for radiographic examination are specified in ASME Section ID, Subarticles NB-5300, NC-5300, and ND-5300. IW A-4221 requires that items used for repair/replacement activities meet the applicable Owner's Requirements and Construction Code requirements when performing repair/replacement activities. IW A-4520 requires that welded joints made for installation of items be examined in accordance with the Construction Code identified in the Repair/Replacement Plan.

4.

Reason for Request

In accordance with 10CFR50.55a(z)(l), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

10 CFR 50.55a Relief Request ISR-05 Revision 0 (Page 2 of 11) Replacement of piping is periodically performed in support of the Flow Accelerated Corrosion (F AC) program as well as other repair and replacement activities. The use of encoded Phased Array Ultrasonic Examination Techniques (PAUT) in lieu of radiography (RT) to perform the required examinations of the replaced welds would eliminate the safety risk associated with performing RT, which includes the planned exposure and the potential for accidental personnel exposure. PAUT minimizes the impact on other outage activities normally involved with performing RT such as limited access to work locations and the need to control system fill status because RT would require a line to remain fluid empty in order to obtain adequate examination sensitivity and resolution. In addition, encoded PAUT has been demonstrated to be adequate for detecting and sizing critical flaws. Exelon Generation Company, LLC (EGC) requests approval of this proposed alternative to support anticipated piping repair and replacement activities for Peach Bottom Atomic Power Station during the fifth ISI interval. S. Proposed Alternative and Basis for Use PBAPS, Unit 2 and 3 is proposing the use of encoded PAUT in lieu of the Code-required RT examinations for Class 1 and 2 ferritic piping repair/replacement welds. Similar techniques are being used throughout the nuclear industry for examination of dissimilar metal welds, and overlaid welds, as well as other applications including ASME B31. l piping replacements. This proposed alternative request includes requirements that provide an acceptable level of quality and safety that satisfy the requirements of 10CFR50.55a(z)(l). The examinations will be performed using personnel and procedures qualified with the requirements of Section 5.1 below. The electronic data files for the PA UT examinations will be stored as part of the arc hi val-quality records. In addition, hard copy prints of the data will also be included as part of the PAUT examination records to allow viewing without the use of hardware or software. 5.1 Proposed Alternative Peach Bottom Atomic Power Station is proposing to perform encoded PAUT examination techniques using demonstrated procedures, equipment, and personnel in accordance with the process documented below: (1) The welds to be examined shall meet the surface conditioning requirements of the demonstrated ultrasonic procedure. (2) The welds to be examined shall be conditioned such that transducers properly couple with the scanning surface with no more than a 1/32 in. (0.8 mm) gap between the search unit and the scanning surface. (3) The ultrasonic examination shall be performed with equipment, procedures, and personnel qualified by performance demonstration.

10 CFR 50.55a Relief Request ISR-05 Revision 0 (Page 3 of 11) ( 4) The examination volume shall include essentially 100% of the weld volume and the weld-to-base-metal interface. (a) Angle beam examination of the complete examination volume for fabrication flaws oriented parallel to the weld joint shall be performed. (b) Angle beam examination for fabrication flaws oriented transverse to the weld joint shall be performed to the extent practical. Scan restrictions that limit complete coverage shall be documented. ( c) A supplemental straight beam examination shall be performed on the volume of base metal through which the angle beams will travel to locate any reflectors that can limit the ability of the angle beam to examine the weld. Detected reflectors that may limit the angle beam examination shall be recorded and evaluated for impact on examination coverage. The straight beam examination procedure, or portion of the procedure, is required to be qualified in accordance with ASME Section V, Article 4 and may be performed using non-encoded techniques. (5) All detected flaw indications from (4)(a) and (4)(b) above shall be considered planar flaws and compared to the preservice acceptance standards for volumetric examination in accordance with IWB-3000, IWC-3000, or IWD-3000. Preservice acceptance standards shall be applied. Analytical evaluation for acceptance of flaws in accordance with IWB-3600, IWC-3600, or IWD-3600 is permitted for flaws that exceed the applicable acceptance standards and are confirmed by surface or volumetric examination to be non-surface connected. (6) Flaws exceeding the applicable acceptance standards and when analytical evaluation has not been performed for acceptance, shall be reduced to an acceptable size or removed and repaired, and the location of the repair shall be reexamined using the same ultrasonic examination procedure that detected the flaw. (7) The ultrasonic examination shall be performed using encoded UT technology that produces an electronic record of the ultrasonic responses indexed to the probe position, permitting off-line analysis of images built from the combined data. (a) Where component configuration does not allow for effective examination for transverse flaws, (e.g., pipe-to-valve, tapered weld transition, weld shrinkage, etc.) the use of non-encoded UT technology may be used for transverse flaws. The basis for the non-encoded examination shall be documented.

10 CFR 50.55a Relief Request ISR-05 Revision 0 (Page 4 of 11) (8) A written ultrasonic examination procedure qualified by performance demonstration shall be used. The qualification shall be applicable to the scope of the procedure, e.g., flaw detection and/or sizing (length or through-wall height), encoded or non-encoded, single and/or dual side access, etc. The procedure shall: (a) contain a statement of scope that specifically defines the limits of procedure applicability (e.g., minimum and maximum thickness, minimum and maximum diameter, scanning access); (b) specify which parameters are considered essential variables, and a single value, a range of values or criteria for selecting each of the essential variables; (c) list the examination equipment, including manufacturer and model or series; (d) define the scanning requirements; such as beam angles, scan patterns, beam direction, maximum scan speed, extent of scanning, and access; (e) contain a description of the calibration method (i.e., actions required to ensure that the sensitivity and accuracy of the signal amplitude and time outputs of the examination system, whether displayed, recorded, or automatically processed, are repeated from examination to examination); (f) describe the method and criteria for discrimination of indications (e.g., geometric indications versus indications of flaws and surface versus subsurface indications); and (g) describe the surface preparation requirements. (9) Performance demonstration specimens shall conform to the following requirements: (a) The specimens shall be fabricated from ferritic material with the same inside surface cladding process, if applicable, with the following exceptions: (i) Demonstration with shielded metal arc weld (SMAW) single-wire cladding is transferable to multiple-wire or strip-clad processes; (ii) Demonstration with multiple-wire or strip-clad process is considered equivalent but is not transferable to SMAW type cladding processes.

10 CFR 50.55a Relief Request 15R-05 Revision 0 (Page 5 of 11) (b) The demonstration specimens shall contain a weld representative of the joint to be ultrasonically examined, including the same welding processes. ( c) The demonstration set shall include specimens not thicker than 0.1 in. (2.5 mm) more than the minimum thickness, nor thinner than 0.5 in. ( 13 mm) less than the maximum thickness for which the examination procedure is applicable. The demonstration set shall include the minimum, within V2 inch of the nominal pipe size (NPS), and maximum pipe diameters for which the examination procedure is applicable. If the procedure is applicable to outside diameter (0.D.) piping of 24 in. (600 mm) or larger, the specimen set must include at least one specimen 24 in. O.D. (600 mm) or larger but need not include the maximum diameter. (d) The demonstration specimen scanning and weld surfaces shall be representative of the surfaces to be examined. (e) The demonstration specimen set shall include geometric conditions that require discrimination from flaws (e.g., counterbore, weld root conditions, or weld crowns) and limited scanning surface conditions for single-side access, when applicable. (f) The demonstration specimens shall include both planar and volumetric fabrication flaws (e.g., lack of fusion, crack, incomplete penetration, slag inclusions) representative of the welding process or processes of the welds to be examined. The flaws shall be distributed throughout the examination volume. (g) Specimens shall be divided into flawed and unflawed grading units. (i) Flawed grading units shall be the actual flaw length, plus a minimum of 0.25 in. (6 mm) on each end of the flaw. Unflawed grading units shall be at least 1 in. (25 mm). (ii) The number of unflawed grading units shall be at least 1-1/2 times the number of flawed grading units. (h) Demonstration specimen set flaw distribution shall be as follows: (i) For thickness greater than 0.50 in. (13 mm); at least 20% of the flaws shall be distributed in the outer third of the specimen wall thickness, at least 20% of the flaws shall be distributed in the middle third of the specimen wall thickness and at least 40% of the flaws shall be distributed

10 CFR 50.55a Relief Request ISR-05 Revision 0 (Page 6 of 11) in the inner third of the specimen wall thickness. For thickness 0.50 in. (13mm) and less, at least 20% of the flaws shall be distributed in the outer half of the specimen wall thickness and at least 40% of the flaws shall be distributed in the inner half of the specimen wall thickness. (ii) At least 30% of the flaws shall be classified as surface planar flaws in accordance with IW A-3310. At least 40% of the flaws shall be classified as subsurface planar flaws in accordance with IW A-3320. (iii) At least 50% of the flaws shall be planar flaws, such as lack of fusion, incomplete penetration, or cracks. At least 20% of the flaws shall be volumetric flaws, such as slag inclusions. (iv) The flaw through-wall heights shall be based on the applicable acceptance standards for volumetric examination in accordance with IWB-3400, IWC-3400, or IWD-3400. At least 30% of the flaws shall be classified as acceptable planar flaws, with the smallest flaws being at least 50% of the maximum allowable size based on the applicable a/I aspect ratio for the flaw. Additional smaller flaws may be included in the specimens to assist in establishing a detection threshold, but shall not be counted as a missed detection if not detected. At least 30% of the flaws shall be classified as unacceptable in accordance with the applicable acceptance standards. Welding fabrication flaws are typically confined to a height of a single weld pass. Flaw through-wall height distribution shall range from approximately one to four weld pass thicknesses, based on the welding process used. (v) If applicable, at least two flaws, but no more than 30% of the flaws, shall be oriented perpendicular to the weld fusion line and the remaining flaws shall be circumferentially oriented. (vi) For demonstration of single-side-access capabilities, at least 30% of the flaws shall be located on the far side of the weld centerline and at least 30% of the planar flaws shall be located on the near side of the weld centerline. The remaining flaws shall be distributed on either side of the weld.

10 CFR 50.55a Relief Request ISR-05 Revision 0 (Page 7 of 11) ( 10) Ultrasonic examination procedures shall be qualified by performance demonstration in accordance with the following requirements. (a) The procedure shall be demonstrated using either a blind or a non-blind demonstration. (b) The non-blind performance demonstration is used to assist in optimizing the examination procedure. When applying the non-blind performance demonstration process, personnel have access to limited knowledge of specimen flaw information during the demonstration process. The non-blind performance demonstration process consists of an initial demonstration without any flaw information, an assessment of the results and feedback on the performance provided to the qualifying candidate. After an assessment of the initial demonstration results, limited flaw information may be shared with the candidate as part of the feedback process to assist in enhancing the examination procedure to improve the procedure performance. In order to maintain the integrity of the specimens for blind personnel demonstrations, only generalities of the flaw information may be provided to the candidate. Procedure modifications or enhancements made to the procedure, based on the feedback process, shall be applied to all applicable specimens based on the scope of the changes. (c) Objective evidence of a flaw's detection, length, and through-wall height sizing, in accordance with the procedure requirements, shall be provided to the organization administering the performance demonstration. (d) The procedure demonstration specimen set shall be representative of the procedure scope and limitations (e.g., thickness range, diameter range, material, access, surface condition). (e) The demonstration set shall include specimens to represent the minimum and maximum diameter and thickness covered by the procedure. If the procedure spans a range of diameters and thicknesses, additional specimens shall be included in the set to demonstrate the effectiveness of the procedure throughout the entire range. (f) The procedure demonstration specimen set shall include at least 30 flaws and shall meet the requirements of (9) above. (g) Procedure performance demonstration acceptance criteria

10 CFR 50.55a Relief Request 15R-05 Revision 0 (Page 8 of 11) (i) To be qualified for flaw detection, all flaws in the demonstration set that are not less than 50% of the maximum allowable size, based on the applicable all aspect ratio for the flaw, shall be detected. In addition, when performing blind procedure demonstrations, no more than 20% of the non-flawed grading units may contain a false call. Any non-flaw condition (e.g., geometry) reported as a flaw shall be considered a false call. (ii) To be qualified for flaw length sizing, the root mean square (RMS) error of the flaw lengths estimated by ultrasonics, as compared with the true lengths, shall not exceed 0.25 in. (6 mm) for diameters of NPS 6.0 in. (DN150) and smaller, and 0.75 in. (18 mm) for diameters greater than NPS 6.0 in. (DN150). (iii) To be qualified for flaw through-wall height sizing, the RMS error of the flaw through-wall heights estimated by ultrasonics, as compared with the true through-wall heights, shall not exceed 0.125 in. (3 mm). (iv) RMS error shall be calculated as follows: RMS= where: mi = measured flaw size n = number of flaws measured ti = true flaw size (h) Essential variables may be changed during successive personnel performance demonstrations. Each examiner need not demonstrate qualification over the entire range of every essential variable. (11) Ultrasonic examination personnel shall be qualified in accordance with IW A-2300. In addition, examination personnel shall demonstrate their capability to detect and size flaws by performance demonstration using the qualified procedure in accordance with the following requirements:

10 CFR 50.55a Relief Request ISR-05 Revision 0 (Page 9 of 11) (a) The personnel performance demonstration shall be conducted in a blind fashion (flaw information is not provided). (i) The demonstration specimen set shall contain at least 10 flaws and shall meet the flaw distribution requirements of (9)(h) above, with the exception of (9)(h)(v). When applicable, at least one flaw, but no more than 20% of the flaws, shall be oriented perpendicular to the weld fusion line and the remaining flaws shall be circumferentially oriented. (b) Personnel performance demonstration acceptance criteria: (i) To be qualified for flaw detection, personnel performance demonstration shall meet the requirements of the following table for both detection and false calls. Any non-flaw condition (e.g., geometry) reported as a flaw shall be considered a false call. Perf onnance Demonstration Detection Test Acceptance Criteria Detection Test Acceptance Criteria False Call Test Acceptance Criteria No. of Flawed Minimum No. of Unflawed Maximum Number of Grading Units Detection Criteria Grading Units False Calls 10 8 15 2 11 9 17 3 12 9 18 3 13 10 20 3 14 10 21 3 15 11 23 3 16 12 24 4 17 12 26 4 18 13 27 4 19 13 29 4 20 14 30 5 Note 1: Flaws ~ 50% of the maximum allowable size, based on the applicable all aspect ratio for the flaw. (ii) To be qualified for flaw length sizing, the RMS error of the flaw lengths estimated by ultrasonics, as compared with the true lengths, shall not exceed 0.25 in. (6 mm) for NPS 6.0 in. (DN150) and smaller, and 0.75 in. (18 mm) for diameters larger than NPS 6.0 in. (DN150).

10 CFR 50.55a Relief Request 15R-05 Revision 0 (Page 10 of 11) (iii) To be qualified for flaw through-wall height sizing, the RMS error of the flaw through-wall heights estimated by ultrasonics, as compared with the true through-wall heights, shall not exceed 0.125 in. (3 mm). (12) Documentation of the qualifications of procedures and personnel shall be maintained. Documentation shall include identification of personnel, NDE procedures, equipment and specimens used during qualification, and results of the performance demonstration. (13) The preservice examinations will be performed per ASME Section XI (Reference 1 ). 5.2 Basis for use The overall basis for this proposed alternative is that encoded PAUT is equivalent or superior to RT for detecting and sizing critical (planar) flaws. In this regard, the basis for the proposed alternative was developed from numerous codes, code cases, associated industry experience, articles, and the results of RT and encoded PAUT examinations. It has been shown that PAUT provides an equally effective examination for identifying the presence of fabrication flaws in carbon steel welds compared to RT (Reference 5). The examination procedure and personnel performing examinations are qualified using representative piping conditions and flaws that demonstrate the ability to detect and size flaws that are both acceptable and unacceptable to the defined acceptance standards. The demonstrated ability of the examination procedure and personnel to appropriately detect and size flaws provides an acceptable level of quality and safety alternative as allowed by 10CFR50.55a(z)(l ). The requirements in this relief request are based upon ASME Section XI Code Case N-831 (N-831) (Reference 3) and will apply to ISi ferritic piping butt welds requiring radiography during repair/replacement activities. N-831 was approved by ASME Board on Nuclear Codes and Standards on October 20, 2016; however, it has not been incorporated into NRC Regulatory Guide 1.14 7, "In service Inspection Code Case Acceptability, ASME Section XI, Division 1," and thus, is not available for application at nuclear power plants without specific NRC approval.

6.

Duration of Proposed Alternative Relief is requested for the Fifth ISi interval for PBAPS, Units 2 and 3, or until the NRC approves N-831, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

7.

Precedents 10 CFR 50.55a Relief Request 15R-05 Revision 0 (Page 11 of 11) PBAPS, Units 2 and 3, fourth ISi interval relief request was authorized by NRC Safety Evaluation (SE) dated June 5, 2017 (Reference 2). This PBAPS relief request was part of an EGC fleet-wide submittal, and the use of encoded phased array ultrasonic examination techniques in lieu of radiography was authorized for various stations. Relief Request 15R-05 for the PBAPS, Units 2 and 3, fifth ISi interval, utilizes a similar approach to the previously approved relief request. Relief Request was authorized for Millstone Power Station, Units 2 and 3, and Surry Power Station, Units 1 and 2; per NRC SE dated January 24, 2018 (Reference 4).

8.

References

1)

ASME Boiler and Pressure Vessel Code, Section XI, 2013 Edition.

2)

Letter from D. J. Wrona (NRC) to B. C. Hanson (EGC), Proposed Alternative to Use Encoded Phased Array Ultrasonic Examination Techniques (CAC Nos. MF8763-MF8782 and MF9395), dated June 5, 2017 (ADAMS Accession No. MLl 7150A091).

3)

ASME Section XI Code Case N-831, "Ultrasonic Examination in Lieu of Radiography for Welds in Ferritic Pipe Section XI, Division 1," ASME Approval Date: October 20, 2016.

4)

Letter from Michael T. Markley (NRC) to Daniel G. Stoddard (Dominion Energy), "Millstone Power Station, Units 2 and 3; and Surry Power Station, Units 1 and 2; Proposed Alternative for the Use of Encoded Phased Array Ultrasonic Examination (CAC Nos. MF9923, MF9924, MF9925, MF9926, MF9927, and MF9928; EPID L-2017-LLR-0060)," dated January 24, 2018 (ADAMS Accession No. ML18019Al95).

5)

US NRC, NUREG/CR-7204, "Applying Ultrasonic Testing in Lieu of Radiography for Volumetric Examination of Carbon Steel Piping" (ADAMS Accession No. ML15253A674).

10 CFR 50.SSa Relief ISR-06 Revision 0 (Page 1 of 13) Request for Reliefs ISR-06 for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange in Accordance with 10 CFR 50.55a(z)(l)

1.

ASME Code Component(s) Affected Code Class:

Reference:

Examination Category: Item Number:

== Description:== Component Number: 1 IWB-2500, Table IWB-2500-1 B-G-1 B6.40 Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, "Threads in Flange" 92 RPV threads in flange for Units 2 and 3 Unit 2 Components: "STUDS 1-46 (THREADS)" and "STUDS 47-92 (THREADS)" Unit 3 Components: "STUDS 1-46 (THREADS)" and "STUDS 47-92 (THREADS)"

2.

Applicable Code Edition and Addenda

The fifth ten-year interval of the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Inservice Inspection (ISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2013 Edition.

3.

Applicable Code Requirement

The Reactor Pressure Vessel (RPV) threads in flange, Examination Category B-G-1, Item Number B6.40, are examined using a volumetric examination technique with 100% of the flange threaded stud holes examined every ISI interval. The examination area is the one-inch area around each RPV stud hole, as shown on Figure IWB-2500-12.

4.

Reason for Request

In accordance with 10CFR50.55a(z)(l), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety. Exelon Generation Company, LLC (EGC) is requesting a proposed alternative from the requirement to perform inservice ultrasonic examinations of Examination Category B-G-1, Item Number B6.40, Threads in Flange for PBAPS, Units 2 and 3. EGC has worked with the industry to evaluate eliminating the RPV threads in flange examination requirement. Licen ees in the U.S. and internationally have worked with

10 CFR 50.55a Relief ISR-06 Revision 0 (Page 2 of 13) the Electric Power Research Institute (EPRI) to produce Technical Report No. 3002007626, "Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements" (Reference 1), which provides the basis for elimination of the requirement. The report includes a survey of inspection results from over 168 units, a review of operating experience related to RPV flange/bolting, and a flaw tolerance evaluation. The conclusion from this evaluation is that the current requirements are not commensurate with the associated burden (worker exposure, personnel safety, radwaste, critical path time, and additional time at reduced water inventory) of the examination. The technical basis for this alternative is discussed in more detail below. Potential Degradation Mechanisms An evaluation of potential degradation mechanisms that could impact flange/threads reliability was performed as part of Reference 1. Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue, stress corrosion cracking, crevice corrosion, velocity phenomena, dealloying corrosion, general corrosion, stress relaxation, creep, mechanical wear, and mechanical/thermal fatigue. Other than the potential for mechanical/thermal fatigue, there are no active degradation mechanisms identified for the threads in flange component. The EPRI report notes a general conclusion from Reference 2, (which includes work supported by the Nuclear Regulatory Commission (NRC)) that when a component item has no active degradation mechanism present, and a preservice inspection has confirmed that the inspection volume is in good condition (i.e., no flaws I indications), then subsequent inservice inspections do not provide additional value going forward. As discussed in the Operating Experience review summary below, the RPV flange ligaments have received the required preservice examinations and over 10,000 inservice inspections, with no relevant findings. To address the potential for mechanical/thermal fatigue, Reference 1 documents a stress analysis and flaw tolerance evaluation of the flange thread area to assess mechanical/thermal fatigue potential. The evaluation consists of two parts. In the first part, a stress analysis is performed considering all applicable loads on the threads in flange component. In the second part, the stresses at the critical locations of the component are used in a fracture mechanics evaluation to determine the allowable flaw size for the component as well as how much time it will take for a postulated initial flaw to grow to the allowable flaw size using guidelines in ASME Section XI, IWB-3500. The Pressurized Water Reactor (PWR) design was selected because of its higher design pressure and temperature. A representative geometry for the finite element model used the largest PWR RPV diameter along with the largest bolts and the highest number of bolts. The larger and more numerous bolt configuration results in less flange material between bolt holes, whereas the larger RPV diameter results in higher pressure and thermal stresses.

Stress Analysis 10 CFR 50.SSa Relief ISR-06 Revision 0 (Page 3 of 13) A stress analysis was performed in Reference 1 to determine the stresses at critical regions of the thread in flange component as input to a flaw tolerance evaluation. Sixteen nuclear plant units (ten PWRs and six Boiling Water Reactors (BWRs)) were considered in the analysis. The evaluation was performed using a geometric configuration that bounds the sixteen units considered in this effort. The details of the RPV parameters for Peach Bottom Atomic Power Station as compared to the values used in the evaluation of the bounding preload stress are shown in Table 1. The preload stresses for both units are bounded by the Reference 1 report. Specifically, the Reference 1 preload stress is 42,338 psi whereas the preload stresses are 30,363 psi at Peach Bottom Atomic Power Station, Units 2 and 3. The Peach Bottom Atomic Power Station stresses are bounded by the Reference 1 report which demonstrates that the report remains applicable to this relief request. For comparison purposes, the global force per flange stud can be estimated by the pressure force on the flange (p*n*r2, where pis the design pressure and r is the vessel inside radius at the stud hole elevation) divided by the number of stud holes. From the parameters in Table 1, this results in a value of 1088 kips per stud for the configuration used in the analysis and 780 kips per stud for the Peach Bottom Atomic Power Station, Units 2 and 3, configurations, indicating that the configuration used in the analysis bounds that at Peach Bottom Atomic Power Station, Units 2 and 3. As shown in Table 1, the preload used in the analysis is also bounding compared to that at Peach Bottom Atomic Power Station, Units 2 and 3. The specifications for the threads and thread geometry for Peach Bottom Atomic Power Station, Units 2 and 3 as compared to that used in the analysis in Reference 1 is shown in Table 2. As this table shows, the flange hole diameter used in the analysis is slightly larger to those at Peach Bottom Atomic Power Station, Units 2 and 3. As can be seen from Table 2, the pitch of the threads used in the analysis is identical to the pitch of the threads for Peach Bottom Atomic Power Station, Units 2 and 3. For Peach Bottom Atomic Power Station, Units 2 and 3, the depths of the thread are slightly smaller than that used in the analysis. However, considering the margins in the analysis for these two plants, these minor differences are considered negligible. Hence the thread geometry used in the analysis is representative of the thread geometry for Peach Bottom Atomic Power Station, Units 2 and 3. Dimensions of the analyzed geometry are shown in Figure 15R-06-1. Table 1: Comparison of Parameters to Values Used in Bounding Analysis No.of Minimum Stud RPV Inside Flange Design Preload Studs Nominal Diameter at Thickness at Plant Currently No. of Studs Diameter Stud Hole Stud Hole Pressure Stress Installed Evaluated (inches) (inches) (inches) (psig) (psi) Peach Bottom Atomic 92 92 6 267.25 14 1280 30,363 Power Station, Unit 2 Peach Bottom Atomic 92 92 6 267.25 14 1280 30,363 Power Station, Unit 3

No.of Studs Plant Currently Installed Values Used in 54 Bounding Analysis 10 CFR 50.55a Relief ISR-06 Revision 0 (Page 4 of 13) Minimum Stud RPV Inside Nominal Diameter at No. of Studs Diameter Stud Hole Evaluated (inches) (inches) 54 6.0 173 Flange Design Pre load Thickness at Stud Hole Pressure Stress (inches) (psig) (psi) 16 2500 42,338 Table 2: RPV Flange Thread Geometry Thread Nominal Bolt Hole Plant Specification Diameter in Flange Pitch Thread Depth (inches) (inches) Peach Bottom Atomic Power 6.75"-8UN-3B 6.75 8 0.06460 Station, Unit 2(1! Peach Bottom Atomic Power 6.75"-8UN-3B 6.75 8 0.06460 Station, Unit 3 Analysis Geometry 7"-8N-2B 7.00 8 0.06500 Note 1: Three out of the 92 bolt holes have Specification 7.00"-8UN-3B with diameter of 7 inches and pitch of 8. The analytical model is shown in Figures I5R-06-2 and I5R-06-3. The loads considered in the analysis consisted of: A design pressure of 2500 psia at an operating temperature of 600°F was applied to all internal surface exposed to internal pressure. Bolt/stud preload - Stress of 42,338 psi. Thermal stresses - The only significant transient affecting the bolting flange is heat-up/cooldown. This transient typically consists of a steady l00°F/hour ramp up to the operating temperature, with a corresponding pressure ramp up to the operating pressure. The ANSYS finite element analysis program was used to determine the stresses in the thread in flange component for the three loads described above. Flaw Tolerance Evaluation A flaw tolerance evaluation was performed using the results of the stress analysis to determine how long it would take an initial postulated flaw to reach the ASME Section XI allowable flaw size. A linear elastic fracture mechanics evaluation consistent with ASME Section XI, IWB-3600 was performed. Stress intensity factors (K's) at four flaw depths of a 360° inside-surface-connected, partial-through-wall circumferential flaws are calculated using finite element analysis techniques with the model described above. The maximum stress intensity factor (K) values around the bolt hole circumference for each flaw depth (a) are extracted and used to perform the crack growth calculations. The circumferential flaw is modeled to start between the 10th and 11th flange threads from the top end of the flange because that is

10 CFR 50.55a Relief ISR-06 Revision 0 (Page 5 of 13) where the largest tensile axial stress occurs. The modeled flaw depth-to-wall thickness ratios (alt) are 0.02, 0.29, 0.55, and 0.77, as measured in any direction from the stud hole. This creates an ellipsoidal flaw shape around the circumference of the flange, as shown in Figure 15R-06-4 for the flaw model with alt= 0.77 alt crack model. The crack tip mesh for the other flaw depths follows the same pattern. When preload is not being applied, the stud, stud threads, and flange threads are not modeled. The model is otherwise unchanged between load cases. The maximum K results are summarized in Table 3 for the four crack depths. From Table 3, the maximum K occurs at operating conditions (preload + heatup +pressure). Because the crack tip varies in depth around the circumference, the maximum K from all locations at each crack size is conservatively used for the K vs. a profile. Table 3: Maximum K vs. alt Load Kat Crack Depth (ksi"in) 0.02 alt 0.29 alt 0.55 alt 0.77 alt Pre load 11.2 17.4 15.5 13.9 Preload + Heatup + Pressure 13.0 19.8 16.1 16.3 The allowable stress intensity factor was determined based on the acceptance criteria in ASME Section XI, IWB-3610/Appendix A which states that: K1 < K1J"l0 = 69.6 ksi"in

Where, K1 =Allowable stress intensity factor (ksi"in)

K1c =Lower bound fracture toughness at operating temperature (220 ksi"in) As can be seen from Table 3, the allowable stress intensity factor is not exceeded for all crack depths up to the deepest analyzed flaw of alt= 0.77. Hence the allowable flaw depth of the 360° circumferential flaw is at least 77% of the thickness of the flange. The allowable flaw depth is assumed to be equal to the deepest modeled crack for the purposes of this analysis. For the crack growth evaluation, an initial postulated flaw size of 0.2 in. (5.08 mm) is chosen consistent with the ASME Section XI, IWB-3500 flaw acceptance standards. The deepest flaw analyzed is alt= 0.77 because of the inherent limits of the model. Two load cases are considered for fatigue crack growth: heat-up/cooldown and bolt preload. The heat-up/cooldown load case includes the stresses due to thermal and internal pressure loads and is conservatively assumed to occur 50 times per year. The bolt preload is assumed to be present and constant during the load cycling of the heat-up/cooldown load case. The bolt preload load case is conservatively assumed to occur five times per year, and these cycles do not include thermal or internal pressure. The resulting crack growth was determined to be negligible due to the small delta Kand the relatively low number of cycles associated with the transients evaluated. Because the crack growth is insignificant,

10 CFR 50.55a Relief ISR-06 Revision 0 (Page 6 of 13) the allowable flaw size will not be reached and the integrity of the component is not challenged for at least 80 years (original 40-year design life plus additional 40 years of plant life extension). An evaluation was also performed to determine the acceptability at preload condition. Table 4 below provides the RPV flange RTNoT values and the bolt-up temperatures for Peach Bottom Atomic Power Station, Units 2 and 3. As this table shows, the information was obtained from plant records as well as the NRC RVID2 database. The values of (T-RTNoT) for the RPV flanges for PBAPS, Units 2 and 3 are shown in Table 4. These were determined using the RTNoT value from plant records. As can be seen from this table, the minimum (T-RTNoT) is 60°F, corresponding to Peach Bottom Atomic Power Station, Units 2 and 3. From the equations in paragraph A-4200 of ASME Section XI, Appendix A, the corresponding value of Kie is 102 ksi~in. Using a structural factor of~ 10, the allowable K1c value is 32.2 ksi~in. This value is more than the maximum stress intensity factor (K1) for the pre load condition of 17.4 ksi ~in shown in Table 3, thus the report evaluation is bounding for PBAPS, Units 2 and 3. Table 4: RPV Flange RTNDT and Bolt-Up Temperature Flange RTNoT (°F) Preload Minimum Plant Name (From Plant Temp (°F) T-RTNDT (°F) Records) Peach Bottom Atomic Power 10 >70 60 Station, Units 2 and 3 The stress analysis I flaw tolerance evaluation presented above shows that the thread in flange component is very flaw tolerant and can operate for 80 years without violating ASME Section XI safety margins. This clearly demonstrates that the thread in flange examinations can be eliminated without affecting the safety of the RPV. Operating Experience Review Summary As discussed above, the results of the survey, which includes results from the Peach Bottom Atomic Power Station, Units 2 and 3, confirmed that the RPV threads in flange examination are adversely impacting outage activities (worker exposure, personnel safety, radwaste, critical path time, and additional time at reduced water inventory) while not identifying any service induced degradations. Specifically, for the U.S. fleet, a total of 94 units responded and none of these units have identified any type of degradation. As can be seen in Table 5 below, the data is encompassing. The 94 units represent data from 33 BWRs and 61 PWRs. For the BWR units, a total 3,793 examinations were conducted and for the PWR units a total of 6,869 examinations were conducted, with no service-induced degradation identified. The response data includes information from all of the plant designs in operation in the U.S. and includes BWR-2, -3, -4, -5, and -6 designs. The PWR plants include the 2-loop, 3-loop, and 4-loop designs and each of the PWR NSSS designs (i.e., Babcock & Wilcox, Combustion Engineering, and Westinghouse).

10 CFR 50.55a Relief 15R-06 Revision 0 (Page 7 of 13) Table 5: Summary of Survey Results - U.S. Fleet Number of Number of Number of Plant Type Units Examinations Reportable Indications BWR 33 3,793 0 PWR 61 6,869 0 Total 94 10,662 0 Related RPV Assessments In addition to the examination history and flaw tolerance discussed above, Reference 1 discusses studies conducted in response to the issuance of the Anticipated Transient Without Scram (ATWS) Rule by the NRC. This rule was issued to require design changes to reduce expected ATWS frequency and consequences. Many studies have been conducted to understand the ATWS phenomena and key contributors to successful response to an ATWS event. In particular, the reactor coolant system (RCS) and its individual components were reviewed to determine weak links. As an example, even though significant structural margin was identified in NRC SECY-83-293 for PWRs, the ASME Service Level C pressure of 3200 psig was assumed to be an unacceptable plant condition. While a higher ASME service level might be defensible for major RCS components, other portions of the RCS could deform to the point of inoperability. Additionally, there was the concern that steam generator tubes might fail before other RCS components, with a resultant bypass of containment. The key take-away for these studies is that the RPV flange ligament was not identified as a weak link and other RCS components were significantly more limiting. Thus, there is substantial structural margin associated with the RPV flange. In summary, Reference 1 identifies that the RPV threads in flange are performing with very high reliability based on operating and examination experience. This is due to the robust design and a relatively benign operating environment (e.g., the number and magnitude of transients is small, generally not in contact with primary water at plant operating temperatures/pressures, etc.) The robust design is manifested in that plant operation has been allowed at several plants even with a bolt/stud assumed to be out of service. As such, significant degradation of multiple bolts/threads would be needed prior to any RCS leakage.

5.

Proposed Alternative and Basis for Use In lieu of the inservice requirements for a ultrasonic examination, Peach Bottom Atomic Power Station, Units 2 and 3 proposes that the industry report (Reference 1) provides an acceptable technical basis for eliminating the requirement for this examination because the alternative maintains an acceptable level of quality and safety.

10 CFR 50.SSa Relief ISR-06 Revision 0 (Page 8 of 13) This report provides the basis for the elimination of the RPV threads in flange examination requirement (ASME Section XI Examination Category B-G-1, Item Number B6.40). This report was developed because evidence had suggested that there have been no occurrences of service-induced degradation and there are negative impacts on worker dose, personnel safety, radwaste, critical path time for these examinations, and additional time at reduced water inventory. Since there is reasonable assurance that the proposed alternative is an acceptable alternate approach to the performance of the ultrasonic examinations, Peach Bottom Atomic Power Station, Units 2 and 3 requests authorization to use the proposed alternative in accordance with 10CFR50.55a(z)(l) on the basis that use of the alternative provides an acceptable level of quality and safety To protect against non-service related degradation, the Peach Bottom Atomic Power Station, Units 2 and 3 uses detailed procedures for the care and visual inspection of the RPV studs and the threads in flange each time the RPV closure head is removed. Care is taken to inspect the RPV threads for damage and to protect threads from damage when the studs are removed. Prior to reinstallation, the studs and stud holes are cleaned and lubricated. The studs are then replaced and tensioned into the RPV flange. This activity is performed each time the closure head is removed, and the procedure documents each step. These controlled maintenance activities provide further assurance that degradation is detected and mitigated prior to returning the reactor to service. The requirements in this relief request are based upon ASME Section XI Code Case N-864 (N-864) (Reference 5) and will apply to Examination Category B-G-1, Item Number B6.40, "Reactor Vessel Threads in Flange." N-864 was approved by ASME Board on Nuclear Codes and Standards on July 28, 2017; however, it has not been incorporated into NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," and thus, is not available for application at nuclear power plants without specific NRC approval.

6.

Duration of Proposed Alternative Relief is requested for the fifth ISI interval for the Peach Bottom Atomic Power Station, Units 2 and 3, or until the NRC approves N-864, or a later revision, in Regulatory Guide 1.14 7 or other document during the interval.

7.

Precedents 10 CFR 50.55a Relief ISR-06 Revision 0 (Page 9 of 13) The Peach Bottom Atomic Power Station, Units 2 and 3, fourth ISi interval relief request was authorized by NRC Safety Evaluation (SE) dated June 26, 2017 (Reference 3). This PBAPS, Units 2 and 3, relief request was part of an EGC fleet-wide submittal, and the alternative for examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, threads in flange was authorized for various stations. Relief Request ISR-06 for the PBAPS, Units 2 and 3, fifth ISi interval, utilizes a similar approach to the previously approved relief request. A relief request was authorized for Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M. Farley Nuclear Plant, Unit 1 by NRC SE dated January 26, 2017 (Reference 4) (ADAMS Accession No. ML17006A109).

8.

References

1)

Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements. EPRI, Palo Alto, CA: 2016. 3002007626 (ADAMS Accession No. ML16221A068).

2)

American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.

3)

Letter from D. J. Wrona (NRC) to B. C. Hanson (EGC) Proposed Alternative to Eliminate Examination of Threads in Reactor Pressure Vessel Flange (CAC Nos. MF8712-MF8729 and MF9548), dated June 26, 2017 (ADAMS Accession No. ML17170A013).

4)

Letter from M. T. Markley (NRC) to C.R. Pierce (Southern Nuclear Operating Co, Inc.) regarding "Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M. Farley Nuclear Plant, Unit 1 - Alternative to lnservice Inspection Regarding Reactor Pressure Vessel Threads Inflange Inspection (CAC Nos. MF8061, MF8062, MF8070)," dated January 26, 2017 (ADAMS Accession No. MLl 7006A109).

5)

ASME Section XI Code Case N-864, "Reactor Vessel Threads in Flange Examination," Section XI, Division 1. ASME Approval Date: July 28, 2017.

R86.5" 17.0" R83.75" R85.69" 10 CFR 50.55a Relief ISR-06 Revision 0 ~ (Page 10 of 13) Figure ISR-06-1 Modeled Dimensions 12.0" 7.0"

• Ir

-~ 16.0" 10.75" I 8.5" / R4. 5"

10 CFR 50.55a Relief ISR-06 Revision 0 (Page 11 of 13) Figure ISR-06-2 Finite Element Model Showing Bolt and Flange Connection lELMNI'S REAL NOM ANO_Vessel_Flange

10 CFR 50.55a Relief ISR-06 Revision 0 (Page 12of13) Figure ISR-06-3 Finite Element Model Mesh with Detail at Thread Location

10 CFR 50.55a Relief ISR-06 Revision 0 (Page 13 of 13) Figure ISR-06-4 Cross Section of Circumferential Flaw with Crack Tip Elements Inserted After 10th Thread from Top of Flange}}