NMP1L3226, Submittal of Relief Requests Associated with the Fifth and Fourth Lnservice Inspection Intervals

Submittal of Relief Requests Associated with the Fifth and Fourth Lnservice Inspection Intervals

ML18159A059
Person / Time
Site:	Nine Mile Point
Issue date:	06/08/2018
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NMP1L3226
Download: ML18159A059 (50)

Text

Exelon Generation ~,

200 Exelon Way Kennett Square. PA 19348 www exeloncorp com 10 CFR 50.55a NMP1L3226 June 8, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRG Docket Nos. 50-220 and 50-410

Subject:

Submittal of Relief Requests Associated with the Fifth and Fourth lnservice Inspection Intervals Attached for your review are relief requests associated with the Fifth and Fourth lnservice Inspection (ISi) intervals for the Nine Mile Point Nuclear Station, Units 1 and 2, respectively.

The fifth and fourth interval programs comply with the 2013 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. The fifth ISi interval at Nine Mile Point Nuclear Station, Unit 1, will begin on August 23, 2019 and is currently scheduled to end August 22, 2029. The fourth ISi interval at Nine Mile Point Nuclear Station, Unit 2, will begin on August 23, 2018 and is currently scheduled to end August 22, 2028.

We request your approval of ISR-04 (NMP1) and 14R-04 (NMP2) concerning Code Case N-513-4 by December 1, 2018.

We request your approval of the remainder of the relief requests by June 8, 2019.

There are no regulatory commitments in this letter.

If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Resp~:: U::--

James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: Relief Requests

Relief Requests Associated with the Fifth and Fourth lnservice Inspection Interval June 8, 2018 Page2 cc: Regional Administrator, Region I, USNRC USNRC Senior Resident Inspector, Nine Mile Point NRC Project Manager, NRR - NMP A. L. Peterson, NYSERDA

Attachment Relief Requests I5R-03 (NMP1) and I4R-03 (NMP2) - BWRVIP Requirements I5R-04 (NMP1) and I4R-04 (NMP2) - N-513-4 I5R-05 (NMP1) and I4R-05 (NMP2) - Threads in Flange I5R-06 (NMP1) and I4R-06 (NMP2) - Use of Encoded Phased Array

10 CFR 50.55a Relief Request Revision 0 (Page 1 of 17)

Request for Reliefs I5R-03 (NMP1) and I4R-03 (NMP2) 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 50.55a(z)(1)

1. ASME Code Component(s) Affected Code Class: 1

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-N-1 and B-N-2 Item Number: B13.10, B13.20, B13.30, and B13.40

Description:

Use of BWRVIP Guidelines in Lieu of Specific ASME Section XI Requirements on the Reactor Pressure Vessel Internals and Components Inspection Component Name: Vessel Interior, Interior Attachments within Beltline Region, Interior Attachments beyond Beltline Region, and Core Support Structure

2. Applicable Code Edition The fifth and fourth 10-year intervals of the Nine Mile Point Nuclear Station, Units 1 and 2, Inservice Inspection (ISI) Programs are 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 Examine accessible areas of the reactor vessel interior each period by the VT-3 visual examination method (B-N-1).

B13.20 Examine interior attachment welds within the beltline region each interval by the VT-1 visual examination method (B-N-2).

B13.30 Examine interior attachment welds beyond the beltline region each interval by the VT-3 visual examination method (B-N-2).

B13.40 Examine accessible surfaces of the core support structure each interval by the VT-3 visual examination method (B-N-2).

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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.

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)(1), 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 Nine Mile Point Nuclear Station, Units 1 and 2 for Examination Categories B-N-1 and B-N-2.

Nine Mile Point Nuclear Station, Units 1 and 2 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

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- BWRVIP-06, Revision 1-A, BWR Vessel and Internals Project, Safety Assessment of BWR Reactor Internals

- 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 4, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines

- BWRVIP-42, Revision 1-A, Low Pressure Coolant Injection System (LPCI) Coupling 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-76, Revision 1-A, BWR Core Shroud Inspection and Flaw Evaluation Guidelines

- BWRVIP-94NP, Revision 2, BWR Vessel and Internals Project, Program Implementation Guide

- BWRVIP-138, Revision 1-A, Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines

- 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.

BWRs 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 with 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 1 through 15.

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 B13.10, 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 Nine Mile Point Nuclear Station, Units 1 and 2.

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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.

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 Nine Mile Point Vessel and Internals February 7, Nuclear Station, ML18040A464 Inspection Summaries for 2018 Unit 1 Spring 2017 Outages (Reference 19)

Project No. 704 - BWR Nine Mile Point Vessel and Internals November 28, Nuclear Station, ML17304A944 Inspection Summaries for 2016 Unit 2* Spring 2016 Outages (Reference 20)

• Note the BWR Vessel and Internals Inspection Summary that includes the latest Nine Mile Point Nuclear Station, Unit 2 outage in Spring 2018 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 Nine Mile Point Nuclear Station, Units 1 and 2 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

10 CFR 50.55a Relief Request Revision 0 (Page 5 of 17) the Executive Committee includes changes to NRC approved BWRVIP inspection 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 requirements of the version of the BWRVIP inspection guideline that forms the safety basis for the NRC-authorized proposed alternative to the requirements of 10 CFR 50.55a, 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.

Nine Mile Point Nuclear Station, Unit 1 is a BWR/2 design and Nine Mile Point Nuclear Station, Unit 2 is a BWR/5 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 Nine Mile Point Nuclear Station, Units 1 and 2.

Therefore, Table 2 only represents the most current comparison. Any deviation 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. Table 3 provides summary information for current deviations applicable to this relief request.

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 IWA-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-R1-A, "BWR Vessel and Internals Project, Safety Assessment of BWR Internals," which has been approved by the NRC. As a result of BWRVIP-06-R1-A, component specific BWRVIP guidelines were developed providing appropriate examination and evaluation requirements to address the specific component safety function and potential degradation mechanism.

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6. Duration of Proposed Alternative Relief is requested for the fifth and fourth ISI intervals for Nine Mile Point Nuclear Station, Units 1 and 2.

7. Precedents

• Relief Request 1ISI-004 and 2ISI-013 was authorized for Nine Mile Point Nuclear Station, Units 1 and 2 fourth and third ISI intervals by NRC SER dated April 29, 2016 (ADAMS Accession No. ML16071A233) (Reference 16).

• Relief Request I4R-02 was authorized conditionally for LaSalle County Generating Station, Units 1 and 2, by NRC SE dated November 17, 2017 (Reference 17).

• 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, December 2009. "Revised Section 4.0 Consideration of Loose Parts of BWRVIP-06-A," BWRVIP letter 2005-207, dated May 2005.

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

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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.

7) "BWRVIP-41, Revision 4: BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002003093, dated July 2014.

8) "BWRVIP-42, Revision 1-A: BWR Vessel and Internals Project, Low Pressure Coolant Injection System (LPCI) Coupling Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 3002010548, dated November 2017.

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

10) 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.

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 T. Tate (NRC) to B. Hanson (EGC), Nine Mile Point Nuclear Station, Units 1 and 2 - Relief Request Alternative RE: Use of Boling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (CAC Nos. MF6116 and MF6117), dated April 29, 2016 (ADAMS Accession No. ML16071A233).

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17) Letter from D. J. Wrona (NRC) B. C. Hanson (EGC), "LaSalle County Station, Units 1 and 2, Relief from the Requirements of the ASME Code and OM Code RE: Relief Requests I4R-02, I4R-03, I4R-06, I4R-07, and I4R-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. ML17305B279).

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 2018-015 from BWRVIP to NRC, "Project No. 704 - BWR Vessel and Internals Inspection Summaries for Spring 2017 Outages," dated February 7, 2018 (ADAMS Accession No. ML18040A464).

20) Letter 2016-132 from BWRVIP to NRC, "Project No. 704 - BWR Vessel and Internals Inspection Summaries for Spring 2016 Outages," dated November 28, 2016 (ADAMS Accession No. ML17304A944).

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TABLE 2 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements1 ASME Item ASME ASME ASME Section XI Authorized BWRVIP Component Section XI Section XI Section XI BWRVIP Exam BWRVIP Frequency No. Table Exam Scope Alternative Exam Scope Exam Frequency IWB-2500-1 B13.10 Reactor Vessel Interior Accessible VT-3 Each period BWRVIP-18-R2-A, Overview examinations of components during BWRVIP examinations Reactor Areas 26-A, 38, 41-R4, are performed to satisfy ASME Section XI VT-3 visual examination Vessel 42-R1-A, 47-A, 48- requirements.

Interior A, 76-R1-A, 138-R1-A, 180, and 183-A B13.20 Jet Pump Riser Braces Accessible VT-1 Each 10- BWRVIP-48-A, Riser Brace EVT-1 25% during each 6 years Interior Welds year Interval Table 3-2 Attachment Attachments Lower Surveillance BWRVIP-48-A, Bracket VT-1 Each 10-year Interval Within Specimen Holder Table 3-2 Attachment Beltline Brackets Region B13.30 Steam Dryer Hold- Accessible VT-3 Each 10- BWRVIP-48-A, Bracket VT-3 Each 10-year Interval Interior Down Brackets Welds year Table 3-2 Attachment Attachments Guide Rod Brackets Interval BWRVIP-48-A, Bracket VT-3 Each 10-year Interval Beyond Table 3-2 Attachment Beltline Steam Dryer Support BWRVIP-48-A, Bracket EVT-1 Each 10-year Interval Brackets Table 3-2 Attachment Feedwater Sparger BWRVIP-48-A, Bracket EVT-1 Each 10-year Interval Brackets Table 3-2 Attachment Core Spray Piping BWRVIP-48-A, Bracket EVT-1 100% every 4 refueling Brackets Table 3-2 Attachment cycles Upper Surveillance BWRVIP-48-A, Bracket VT-3 Each 10-year Interval Specimen Holder Table 3-2 Attachment Brackets Shroud Support Welds BWRVIP-38, Weld H93 EVT-1 or UT Based on as-found 3.3, Figure 3-5 conditions, to a maximum

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TABLE 2 Comparison of ASME Section XI Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements1 ASME Item ASME ASME ASME Section XI Authorized BWRVIP Component Section XI Section XI Section XI BWRVIP Exam BWRVIP Frequency No. Table Exam Scope Alternative Exam Scope Exam Frequency IWB-2500-1 of 6 years for EVT-1, 10 years for UT Shroud Support Leg Accessible VT-3 Each 10- BWRVIP-38, Weld H12 Per BWRVIP-38 When accessible (Weld H12) Welds year 3.2.3 NRC SE (Beneath core Interval (07/24/00),

plate, rarely accessible) examine with appropriate method4 B13.40 Shroud Support2 Accessible VT-3 Each BWRVIP-38, Welds H8 and EVT-1 or UT Based on as-found Core Support Surfaces 10-year 3.3, H93 conditions, to a maximum of Structure Interval Appendix A 6 years for EVT-1, 10 years Figures 3-4 and 3-5 for UT Shroud Support Legs Accessible BWRVIP-38, Shroud support Per BWRVIP-38 When accessible (NMP2 only) Surfaces 3.2.3 leg welds NRC SE (beneath core (07/24/00),

plate, rarely accessible) examine with appropriate method4 Shroud Horizontal Accessible BWRVIP-76-R1-A, Welds H1 - H7 EVT-1 or UT Based on as-found Welds Surfaces 2.2, Figure 2-3 as applicable conditions, to a maximum (NMP2 only) of 6 years for one-sided EVT-1, 10 years for UT Shroud Vertical Welds BWRVIP-76-R1-A, Vertical and EVT-1 or UT Maximum of 6 years for 2.3, 3.3, Figures 2- Ring Segment one-sided EVT-1, 10 years 4, 2-5, and 3-2 Welds for UT Shroud Repairs5 Accessible BWRVIP-76-R1-A, Tie-Rod Repair VT-3 Per repair designer (NMP1 only) Surfaces 3.5 and 3.6 recommendations, per BWRVIP-76-R1-A

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NOTES:

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) The NMP1 is a BWR/2 design and has a conical shroud support plate. The NMP2 design is a shroud support plate with legs.

3) The NMP1 evaluation in Appendix A of BWRVIP-38 identifies the inspection scope based on inspection frequency. Weld H9 inspection evaluation and frequency was previously accepted by the NRC in a letter dated October 31, 2001 (ADAMS Accession No. ML012990403).

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

5) NMP1 has a tie rod shroud repair. NMP2 does not have a shroud repair.

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TABLE 3 BWRVIP Deviations BWRVIP PLANT LETTER DATE TO USNRC DEVIATION APPLICABILITY DOCUMENT Nine Mile Point BWRVIP-25 Letter from Joseph E. Pacher BWRVIP-25 requires ultrasonic or visual This deviation does Nuclear Station, Unit (Constellation Energy, LLC) to examination of the core plate bolts. There are not impact the basis 2 U.S. Nuclear Regulatory currently no examination methods available for for use of this relief Commission, dated March 30, these bolts and is being addressed as a BWRVIP request.

2011 generic issue. Analytical evaluation has been completed to support operation.

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Enclosure 1 Comparison of ASME Section XI 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 10-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), top guide (BWRVIP-26-A), shroud support (BWRVIP-38), jet pump welds and components (BWRVIP-41-R4), LPCI couplings (BWRVIP-42-R1-A), lower plenum components (BWRVIP-47-A) interior attachments (BWRVIP-48-A), core shroud welds, (BWRVIP-76-R1-A), access hole cover (BWRVIP-180), and top guide grid beams (BWRVIP-183-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

10 CFR 50.55a Relief Request Revision 0 (Page 14 of 17) examination of the interior of the reactor vessel. Accordingly, these BWRVIP 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 (B-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 (GE) Company BWR/2 and BWR/5 designs, 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 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.

• BWRVIP-48-A requires an EVT-1 visual examination of the jet pump riser brace-to-reactor vessel wall pad welds and heat affected zones 25% each 6 years.

The 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 and cleaning as necessary. BWRVIP-48-A includes a diagram for the configuration and prescribes examination for each configuration including Nine Mile Point Nuclear Station, Unit 2.

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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/2 and BWR/5 designs, this includes the core spray piping 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 welds-to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, and the shroud support plate-to-reactor vessel weld. 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 XI.

For the core spray support bracket 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 feedwater sparger 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 - Feedwater Sparger Bracket Welds (BWRVIP-48-A)

• The ASME Section XI examination requirement is a VT-3 visual examination of each weld every 10 years.

• The BWRVIP-48-A visual examination requirement is an EVT-1 for the feedwater sparger bracket attachment welds with each weld examined every 10 years.

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The BWRVIP-48-A visual examination method EVT-1 has superior flaw detection and sizing capability, the examination frequency is the same as the ASME Section XI requirements, and the same flaw evaluation criteria are used.

An 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, and the relevant degradation mechanisms for BWR internal attachments.

Therefore, because the EVT-1 visual examination method provides the same examination scope (accessible welds), the same examination frequency in most cases, and the same flaw evaluation criteria as ASME Section XI, the level of quality and safety provided by the BWRVIP criteria meets or 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/2 and BWR/5 designs, the core support structure has primarily been considered the shroud support structure, including the shroud. 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-1 visual examination every 6 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 UT of the susceptible welds at a frequency identical to the ASME Section XI requirement.

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.

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Comparison to BWRVIP Requirements - Core Shroud (BWRVIP-76-R1-A)

Shroud repair tie-rods have been installed at Nine Mile Point Nuclear Station, Unit 1.

Therefore, BWRVIP-76-R1-A requires inspection of the vertical shroud welds and the tie-rod repair hardware.

• ASME Section XI requires a VT-3 visual examination of accessible surfaces each 10-year interval.

• The BWRVIP-76-R1-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.

• For shroud repairs, the BWRVIP requires either an EVT-1 visual examination or UT of shroud vertical welds every 10-years minimum, as compared to the ASME Section XI requirement (VT-3).

• For shroud repairs, the BWRVIP requires a VT-3 visual examination and other appropriate techniques to examine the tie-rod repair hardware every ten years.

Therefore, the BWRVIP referenced examinations are the same or superior to ASME Section XI requirements. Shroud vertical welds and repair tie-rod examinations are recommended in BWRVIP-76-R1-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 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 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.

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Request for Reliefs I5R-04 (NMP1) and I4R-04 (NMP2) to Use ASME Code Case N-513-4 In Accordance with 10 CFR 50.55a(z)(2)

1. ASME Code Component(s) Affected All American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV),Section XI, Class 2 and 3 components that meet the operational and configuration limitations of ASME Code Case N-513-4 (N-513-4), "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping,Section XI, Division 1," paragraphs 1(a), 1(b), 1(c), and 1(d).

2. Applicable Code Edition The fifth and fourth 10-year intervals of the Nine Mile Point Nuclear Station, Units 1 and 2, Inservice Inspection (ISI) Programs are based on the ASME B&PV Code,Section XI, 2013 Edition.

3. Applicable Code Requirement IWC-3120 and IWD-3120 of ASME Section XI, require that flaws exceeding the defined acceptance criteria be corrected by repair/replacement activities or evaluated and accepted by analytical evaluation. IWC-3130 and IWD-3130 of ASME Section XI, require that relevant conditions be subject to supplemental examination, corrective measures or repair/replacement activities, or evaluated and accepted by analytical evaluation.

4. Reason for Request 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.

Exelon Generation Company, LLC (EGC) is requesting a proposed alternative from the requirement to perform repair/replacement activities for degraded Class 2 and 3 piping whose maximum operating temperature does not exceed 200°F and whose maximum operating pressure does not exceed 275 psig for Nine Mile Point Nuclear Station, Units 1 and 2. Moderately degraded Class 2 and 3 piping could require a plant shutdown within the required action statement timeframes to repair observed degradation. Plant shutdown activities result in additional dose and plant risk that would be inappropriate when a degraded condition is demonstrated to retain adequate margin to complete the component's function. The use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition will allow EGC to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long term repair actions if necessary. Actions to remove degraded piping from

10 CFR 50.55a Relief Request Revision 0 (Page 2 of 5) service could have a detrimental overall risk impact by requiring a plant shutdown, thus requiring use of a system that is in standby during normal operation. Accordingly, compliance with the current Code requirements results in a hardship without a compensating increase in the level of quality and safety.

ASME Code Case N-513-3 (N-513-3) does not allow evaluation of flaws located away from attaching circumferential piping welds that are in elbows, bent pipe, reducers, expanders, and branch tees. N-513-3 also does not allow evaluation of flaws located in heat exchanger external tubing or piping. N-513-4 provides guidance for evaluation of flaws in these locations.

5. Proposed Alternative and Basis for Use EGC is requesting approval to apply the evaluation methods of N-513-4 to Class 2 and 3 components that meet the operational and configuration limitations of N-513-4, paragraphs 1(a), 1(b), 1(c), and 1(d) for Nine Mile Point Nuclear Station, Units 1 and 2 in order to avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with the cited Code requirements.

The Nuclear Regulatory Commission (NRC) issued Generic Letter 90-05 (Reference 1),

"Guidance for Performing Temporary Non-Code Repair of Class 1, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptability of limited degradation in moderate energy piping. The generic letter defines conditions that would be acceptable to utilize temporary non-Code repairs with NRC approval. The ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed ASME Code Case N-513 (N-513). NRC approval of N-513 versions in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 18 (Reference 4), allows acceptance of partial through-wall or through-wall leaks for an operating cycle provided all conditions of the code case and NRC conditions are met. The code case also requires the Owner to demonstrate system operability due to leakage.

The ASME recognized that the limitations in N-513-3 were preventing needed use in piping components such as elbows, bent pipe, reducers, expanders, and branch tees and external tubing or piping attached to heat exchangers. N-513-4 was approved by the ASME to expand use on these locations and to revise several other areas of the code case.

Attachment 2 of the Reference 2 letter provides a marked-up N-513-3 version of the code case to highlight the changes compared to the NRC approved N-513-3 version.

Attachment 3 of the Reference 2 letter provides the ASME approved N-513-4. The following provides a high level overview of the N-513-4 changes:

1) Revised the maximum allowed time of use from no longer than 26 months to the next scheduled refueling outage.

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2) Added applicability to piping elbows, bent pipe, reducers, expanders, and branch tees where the flaw is located more than (Rot)1/2 from the centerline of the attaching circumferential piping weld.

3) Expanded use to external tubing or piping attached to heat exchangers.

4) Revised to limit the use to liquid systems.

5) Revised to clarify treatment of Service Level load combinations.

6) Revised to address treatment of flaws in austenitic pipe flux welds.

7) Revised to require minimum wall thickness acceptance criteria to consider longitudinal stress in addition to hoop stress.

8) Other minor editorial changes to improve the clarity of the code case.

Detailed discussion of significant changes in N-513-4 when compared to NRC approved N-513-3 is provided in Attachment 4 of the Reference 2 letter.

The design basis is considered for each leak and evaluated using the EGC Operability Evaluation process. The evaluation process must consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgement. As required by the code case, the evaluation process considers but is not limited to system make-up capacity, containment integrity with the leak not isolated, effects on adjacent equipment, and the potential for room flooding.

Leakage rate is not typically a good indicator of overall structural stability in moderate energy systems, where the allowable through-wall flaw sizes are often on the order of inches. The periodic inspection interval defined using paragraph 2(e) of N-513-4 provides evidence that a leaking flaw continues to meet the flaw acceptance criteria and that the flaw growth rate is such that the flaw will not grow to an unacceptable size.

The effects of leakage may impact the operability determination or the plant flooding analyses specified in paragraph 1(f). For a leaking flaw, the allowable leakage rate will be determined by dividing the critical leakage rate by a safety factor of four (4). The critical leakage rate is determined as the lowest leakage rate that can be tolerated and may be based on the allowable loss of inventory or the maximum leakage than can be tolerated relative to room flooding, among others. The safety factor of four (4) on leakage is based upon ASME Code Case N-705 (N-705) (Reference 3), which is accepted without condition in Regulatory Guide 1.147, Revision 18. Paragraph 2.2(e) of N-705 requires a safety factor of two (2) on flaw size when estimating the flaw size from the leakage rate. This corresponds to a safety factor of four (4) on leakage for nonplanar flaws. Although the use of a safety factor for determination of an unknown flaw is considered conservative when the actual flaw size is known, this approach is deemed acceptable based upon the precedent of N-705. Note that the alternative herein does not

10 CFR 50.55a Relief Request Revision 0 (Page 4 of 5) propose to use any portion of N-705 and that citation of N-705 is intended only to provide technical basis for the safety factor on leakage.

During the temporary acceptance period, leaking flaws will be monitored daily as required by paragraph 2(f) of N-513-4 to confirm the analysis conditions used in the evaluation remain valid. Significant change in the leakage rate is reason to question that the analysis conditions remain valid, and would require re-inspection per paragraph 2(f) of the code case. Any re-inspection must be performed in accordance with paragraph 2(a) of the code case.

The leakage limit provides quantitative measurable limits which ensure the operability of the system and early identification of issues that could erode defense-in-depth and lead to adverse consequences.

In summary, EGC will apply N-513-4 to the evaluation of Class 2 and 3 components that are within the scope of the code case. N-513-4 utilizes technical evaluation approaches that are based on principals that are accepted in other Code documents already acceptable to the NRC. The application of this code case, in conjunction with safety factors on leakage limits, will maintain acceptable structural and leakage integrity while minimizing plant risk and personnel exposure by minimizing the number of plant transients that could be incurred if degradation is required to be repaired based on ASME Section XI acceptance criteria only.

6. Duration of Proposed Alternative The proposed alternative is for use of N-513-4 for Class 2 and Class 3 components within the scope of the code case. An ASME Section XI compliant repair/replacement will be completed prior to exceeding the next refueling outage or allowable flaw size, whichever comes first. Relief is requested for the fifth and fourth ISI intervals for Nine Mile Point Nuclear Station, Units 1 and 2, or until the NRC approves N-513-4, or a later revision, in Regulatory Guide 1.147 or other document during the interval. If a flaw is evaluated near the end of the interval for Nine Mile Point Nuclear Station, Units 1 and 2 and the next refueling outage is in the subsequent interval, the flaw may remain in service under this relief request until the next refueling outage.

7. Precedents Nine Mile Point Nuclear Station, Units 1 and 2, fourth and third ISI intervals relief request was authorized by NRC Safety Evaluation (SE) dated September 6, 2016 (Reference 5). This Nine Mile Point Nuclear Station, Units 1 and 2 relief request was part of an EGC fleet-wide submittal, and the alternative for the use of N-513-4 was authorized for various stations. This relief request for the Nine Mile Point Nuclear Station, Units 1 and 2, fifth and fourth ISI intervals, utilizes a similar approach to the previously approved relief request.

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8. References

1) NRC Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," dated June 15, 1990.

2) Letter from D. T. Gudger (Exelon Generation Company, LLC) to NRC, "Proposed Alternative to Utilize Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1," dated January 28, 2016.

3) ASME Section XI Code Case N-705, "Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and TanksSection XI, Division 1," dated October 12, 2006.

4) NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 18, dated February 16, 2018.

5) Letter from G. Miller (U.S. Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), Proposed Alternative to Use ASME Code Case N-513-4, dated September 6, 2016 (CAC NOS. MF7301-MF7322)

(ADAMS Accession No. ML16230A237).

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Request for Reliefs I5R-05 (NMP1) and I4R-05 (NMP2) 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)(1)

1. ASME Code Component(s) Affected Code Class: 1

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-G-1 Item Number: B6.40

Description:

Alternative for Examination of ASME Section XI, Examination Category B-G-1, Item Number B6.40, Threads in Flange Component Number: 64 RPV threads in flange for Units 1 (RV-01-L to RV L) 76 RPV threads in flange for Units 2 (2RPV-TF001 to 2RPV-TF076)

2. Applicable Code Edition The fifth and fourth 10-year intervals of the Nine Mile Point Nuclear Station, Units 1 and 2, Inservice Inspection (ISI) Programs are 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 10 CFR 50.55a(z)(1), 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 Nine Mile Point Nuclear Station, Units 1 and 2. EGC has worked with the industry to evaluate eliminating the RPV threads in flange examination requirement. Licensees in the U.S. and internationally have worked with 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

10 CFR 50.55a Relief Request Revision 0 (Page 2 of 13) 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 / 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 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

10 CFR 50.55a Relief Request Revision 0 (Page 3 of 13) 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 Nine Mile Point Nuclear Station, Units 1 and 2 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 25,356 psi and 27,411 psi at Nine Mile Point Nuclear Station, Units 1 and 2. The Nine Mile Point Nuclear 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**r2, where p is 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 707 kips (Unit 1) and 827 kips (Unit 2) per stud for the Nine Mile Point Nuclear Station, Units 1 and 2 configurations, indicating that the configuration used in the analysis bounds that at Nine Mile Point Nuclear Station, Units 1 and 2. As shown in Table 1, the preload stress used in the analysis is also bounding compared to that at Nine Mile Point Nuclear Station, Units 1 and 2.

The specifications for the threads and thread geometry for Nine Mile Point Nuclear Station, Units 1 and 2 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 equal to or smaller than those at Nine Mile Point Nuclear Station, Units 1 and 2. The larger hole diameter results in a smaller remaining ligament between holes, and is therefore conservative. 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 Nine Mile Point Nuclear Station, Units 1 and 2.

For Nine Mile Point Nuclear Station, Unit 1, the depth of the thread is slightly smaller than that used in the analysis. However, considering the margins in the analysis for these two plants, this minor difference is considered negligible. Hence the thread geometry used in the analysis is representative of the thread geometry for Nine Mile Point Nuclear Station, Units 1 and 2. Dimensions of the analyzed geometry are shown in Figure I5R-05-1/I4R-05-1.

Table 1: Comparison of Parameters to Values Used in Bounding Analysis No. of Minimum Stud RPV Inside Flange Design Studs No. of Nominal Diameter at Thickness at Preload Plant Pressure Currently Studs Diameter Stud Hole Stud Hole Stress (psi)

(psig)

Installed Evaluated (inches) (inches) (inches)

Nine Mile Point, Unit 1 64 64 6.25 213.44 13.84 1265 25,356 Nine Mile Point, Unit 2 76 76 6.5 251.5 13.5 1265 27,411 Values Used in 54 54 6.0 173 16 2500 42,338 Bounding Analysis

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Table 2: RPV Flange Thread Geometry Nominal Bolt Hole Thread Thread Plant Diameter in Flange Pitch Depth Specification (inches) (inches)

Nine Mile Point, Unit 1 7"-8-2B Special 7.00 8 0.06345 Nine Mile Point, Unit 2 6.25"-8UN-2B 6.25 8 0.06700 Analysis Geometry 7"-8N-2B 7.00 8 0.06500 The analytical model is shown in Figures I5R-05-2/I4R-05-2 and I5R-05-3/I4R-05-3.

The loads considered in the analysis consisted of:

• A design pressure of 2500 psia at an operating temperature of 600oF 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 100ºF/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 where the largest tensile axial stress occurs. The modeled flaw depth-to-wall thickness ratios (a/t) 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 I4R-05-4 for the flaw model with a/t = 0.77 a/t 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).

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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. a/t K at Crack Depth (ksiin)

Load 0.02 a/t 0.29 a/t 0.55 a/t 0.77 a/t Preload 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:

KI < KIc/10 = 69.6 ksiin

Where, KI = Allowable stress intensity factor (ksiin)

KIc = Lower bound fracture toughness at operating temperature (220 ksiin)

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 a/t = 0.77. Hence the allowable flaw depth of the 360o 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 a/t = 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 K and the relatively low number of cycles associated with the transients evaluated. Because the crack growth is insignificant, 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 RTNDT values and the bolt-up temperatures for Nine Mile Point Nuclear Station, Units 1 and 2. These were determined using the RTNDT value from plant records. As can be seen from this table, the minimum (T-RTNDT) is 10oF and 60oF, corresponding to Nine Mile Point Nuclear Station, Units 1 and 2, respectively.

From the equations in paragraph A-4200 of ASME Section XI, Appendix A, the

10 CFR 50.55a Relief Request Revision 0 (Page 6 of 13) corresponding values of KIc are 59 and 102 ksiin. Using a structural factor of 10, the allowable KIc value is 18.5 and 32.3 ksiin. This value is more than the maximum stress intensity factor (KI) for the preload condition of 17.4 ksiin shown in Table 3, thus the report evaluation is bounding for Nine Mile Point Nuclear Station, Units 1 and 2.

Table 4: RPV Flange RTNDT and Bolt-Up Temperature Minimum Plant Name Flange RTNDT (oF) Preload Temp (oF)

T-RTNDT (oF)

Nine Mile Point, Unit 1 60 70 to 120 10 Nine Mile Point, Unit 2 10 >70 60 The stress analysis / 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 Nine Mile Point Nuclear Station, Units 1 and 2, 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 have responded to date 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).

Table 5: Summary of Survey Results - U.S. Fleet Number of Number of Number of Plant Type Reportable Units Examinations Indications BWR 33 3,793 0 PWR 61 6,869 0 Total 94 10,662 0

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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 volumetric ultrasonic examination, Nine Mile Point Nuclear Station, Units 1 and 2 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.

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, Nine Mile Point Nuclear Station, Units 1 and 2 requests authorization to use the proposed alternative in accordance

10 CFR 50.55a Relief Request Revision 0 (Page 8 of 13) with 10CFR50.55a(z)(1) on the basis that use of the alternative provides an acceptable level of quality and safety.

To protect against non-service related degradation, Nine Mile Point Nuclear Station, Units 1 and 2 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 and fourth ISI intervals for Nine Mile Point Nuclear Station, Units 1 and 2, or until the NRC approves N-864, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

7. Precedents

• Nine Mile Point Nuclear Station, Units 1 and 2, fourth and third ISI intervals relief request was authorized by NRC Safety Evaluation (SE) dated June 26, 2017 (Reference 3). This Nine Mile Point Nuclear Station, Units 1 and 2 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. This relief request for the Nine Mile Point Nuclear Station, Units 1 and 2, fifth and fourth ISI intervals, utilizes a similar approach to the previously approved relief request.

• 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).

10 CFR 50.55a Relief Request Revision 0 (Page 9 of 13)

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 Inservice Inspection Regarding Reactor Pressure Vessel Threads in Flange Inspection (CAC Nos. MF8061, MF8062, MF8070)," dated January 26, 2017 (ADAMS Accession No. ML17006A109).

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

10 CFR 50.55a Relief Request Revision 0 (Page 10 of 13)

Figure I5R-05-1/I4R-05-1 Modeled Dimensions R86.5 8.5 12.0 17.0 7.0 16.0 R83.75 R4.5 10.75 R85.69

10 CFR 50.55a Relief Request Revision 0 (Page 11 of 13)

Figure I5R-05-2/I4R-05-2 Finite Element Model Showing Bolt and Flange Connection

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Figure I5R-05-3/I4R-05-3 Finite Element Model Mesh with Detail at Thread Location

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Figure I5R-05-4/I4R-05-4 Cross Section of Circumferential Flaw with Crack Tip Elements Inserted After 10th Thread from Top of Flange

10 CFR 50.55a Relief Request Revision 0 (Page 1 of 12)

Request for Reliefs I5R-06 (NMP1) and I4R-06 (NMP2) for the Use of Encoded Phased Array Ultrasonic Examination Techniques in Lieu of Radiography In Accordance with 10 CFR 50.55a(z)(1)

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

2. Applicable Code Edition The fifth and fourth 10-year intervals of the Nine Mile Point Nuclear Station, Units 1 and 2, Inservice Inspection (ISI) Programs are based on the ASME B&PV Code,Section XI, 2013 Edition.

3. Applicable Code Requirement 10CFR50.55a(b)(2)(xx)(B) requires that "The NDE provision in IWA-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)(1)(ii) of this section."

IWA-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 III 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 III, Subarticles NB-5200, NC-5200, and ND-5200. The acceptance standards for radiographic examination are specified in ASME Section III, Subarticles NB-5300, NC-5300, and ND-5300.

IWA-4221 requires that items used for repair/replacement activities meet the applicable Owner's Requirements and Construction Code requirements when performing repair/replacement activities. IWA-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)(1), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

Replacement of piping is periodically performed in support of the Flow Accelerated Corrosion (FAC) program as well as other repair and replacement activities. The use of

10 CFR 50.55a Relief Request Revision 0 (Page 2 of 12) 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 Nine Mile Point Nuclear Station, Units 1 and 2 during the fifth and fourth ISI intervals.

5. Proposed Alternative and Basis for Use Nine Mile Point Nuclear Station, Units 1 and 2 is proposing the use of encoded PAUT in lieu of the Code-required RT examinations for ISI 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 B31.1 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)(1). The examinations will be performed using personnel and procedures qualified with the requirements of Section 5.1 below.

The electronic data files for the PAUT examinations will be stored as part of the archival-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 Nine Mile Point Nuclear 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 Revision 0 (Page 3 of 12)

(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 Revision 0 (Page 4 of 12)

(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 Revision 0 (Page 5 of 12)

(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 1/2 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 (O.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 Revision 0 (Page 6 of 12) 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 IWA-3310. At least 40%

of the flaws shall be classified as subsurface planar flaws in accordance with IWA-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 Revision 0 (Page 7 of 12)

(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.

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(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 a/I 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:

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 IWA-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:

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(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.

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Performance 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 a/ 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).

(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

10 CFR 50.55a Relief Request Revision 0 (Page 11 of 12) 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)(1).

The requirements in this relief request are based upon ASME Section XI Code Case N-831 (N-831) (Reference 4) 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 1.147, "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 and fourth ISI intervals for Nine Mile Point Nuclear Station, Units 1 and 2, or until the NRC approves N-831, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

7. Precedents

• Nine Mile Point Nuclear Station, Units 1 and 2, fourth and third ISI intervals relief request was authorized by NRC Safety Evaluation (SE) dated June 5, 2017 (Reference 2). This Nine Mile Point Nuclear Station 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.

This relief request for the Nine Mile Point Nuclear Station, Units 1 and 2, fifth and fourth ISI intervals, utilizes a similar approach to the previously approved relief request.

• Relief request was authorized for Millstone Power Station, Units 1 and 2, and Surry Power Station, Units 1 and 2 by NRC SE dated January 24, 2018 (Reference 3).

8. References

1) ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components," 2013 Edition.

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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. ML17150A091).

3) Letter from Michael T. Markley, US NRC, to Daniel G. Stoddard, Dominion Energy,

Subject:

Millstone Power Station, Units 1 and 2, 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. ML18019A195).

4) 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.

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