ML19308B926

ACRS Plant License Renewal Subcommittee - Peach Bottom Atomic Power Station Units 2 and 3 Subsequent License Renewal Safety Evaluation Report (SER) with Confirmatory Item

ML19308B926
Person / Time
Site:	Peach Bottom
Issue date:	11/05/2019
From:	Bennett Brady NRC/NRR/DNRL/NLRP
To:
Burton B, NRR-DNRL 415-2981
References
Download: ML19308B926 (23)

Text

Advisory Committee on Reactor Safeguards Plant License Renewal Subcommittee Peach Bottom Atomic Power Station Units 2 and 3 Subsequent License Renewal Safety Evaluation Report (SER) with Confirmatory Item November 5, 2019 Bennett Brady, Senior Project Manager Office of Nuclear Reactor Regulation

Presentation Outline

• Overview of Safety Review of Peach Bottom SLRA

• SER Section 2, Scoping and Screening Review

• SER Section 3, Aging Management Review

• SER Section 4, Time-Limited Aging Analyses

• Closure of Confirmatory Item

• Specific Areas of Review - Reactor Vessel and Internals, Irradiated Concrete, and Electrical Cables

• SLRA Review Conclusion

• Region Presentation on Inspections and Plant Material Conditions

• Region Conclusion

• Summary Conclusion 2

Overview of Safety Review of Peach Bottom SLRA Unit Initial Initial License Renewed Expiration Subsequent License License Renewal License Date Renewal Application Application 2 10/25/1973 07/02/2001 05/07/2003 08/08/2033 07/10/2018 3 07/02/1974 07/02/2001 05/07/2003 07/02/2034 07/10/2018

• Application Submitted - July 10, 2018

• Acceptance Determination - September 6, 2018

• Safety Evaluation Report with Confirmatory Item -

October 7, 2019 3

Audits and Inspections Dates Location Operating September 17-27, Rockville, MD Experience Audit 2018 November 13, In-office Audit 2018 - April 29, Rockville, MD 2019 4

SER Overview

• SER with Confirmatory Item Issued October 7, 2019

- Confirmatory Item (CI) 3.0.3.2.3-1 BWR Vessel Internals

• Requests for Additional Information (RAIs)

- 48 RAIs issued, 4 of which were follow-up RAIs 5

SER Section 2 Structures and Components Subject to Aging Management Review (AMR)

• Section 2.1 Scoping and Screening Methodology

• Section 2.2 Plant Level Scoping Results

• Sections 2.3, 2.4, 2.5, Scoping and Screening Results 6

SER Section 3 Aging Management Review (AMR)

• Section 3.0 Use of the Generic Aging Lessons Learned Report

• Section 3.1 Reactor Vessel, Internals, and Reactor Coolant System

• Section 3.2 Engineered Safety Features

• Section 3.3 Auxiliary Systems

• Section 3.4 Steam and Power Conversion Systems

• Section 3.5 Containment, Structures and Component Supports

• Section 3.6 Electrical and Instrumentation and Control Commodities 7

SER Section 3 3.0.3 - Aging Management Programs (AMPs)

SLRA - Original Disposition of AMPs SER - Final Disposition of AMPs

• 11 new GALL programs

• 11 new GALL programs 8 consistent 8 consistent 3 consistent with 3 consistent with enhancements enhancements

• 35 existing GALL programs

• 35 existing GALL programs

- 8 consistent - 8 consistent

- 27 consistent with - 27 consistent with enhancements/exceptions enhancements/exceptions

• 1 plant-specific with

• 1 plant-specific with enhancement enhancement evaluated against Appendix A1 of the SRP-SLR 8

SER Section 4 Time-Limited Aging Analyses (TLAAs)

• 4.1 Identification of TLAAs

• 4.2 Reactor Vessel and Internals Neutron Embrittlement Analyses

• 4.3 Metal Fatigue Analyses

• 4.4 Environmental Qualification of Electric Equipment

• 4.5 Concrete Containment Tendon Prestress Analysis

• 4.6 Primary Containment Fatigue Analysis

• 4.7 Other Plant-Specific TLAAs 9

Confirmatory Item 3.0.3.2.3-1 BWR Vessel Internals Issue: SLRA, AMP B.2.1.7 BWR Vessel Internals proposed to either:

• install core plate wedges or

• submit for NRC approval an inspection plan for the core plate rim hold-down bolts to mitigate stress corrosion cracking.

Staff determined the second option - a future inspection plan - does not provide sufficient information for the staffs evaluation.

Resolution: Applicant enhances AMP B.2.1.7, in accordance with BWRVIP-25, Revision 1, to:

• install wedges or

• inspect core plate rim hold-down bolts, or

• demonstrate instead via analysis that the installation of wedges and inspections of the core plate rim hold-down bolts are not required.

The staff determined each of the three proposed options was acceptable.

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Specific Areas of SLRA Review

• Reactor Pressure Vessel Neutron Embrittlement

• Reactor Vessel Internals - Irradiation-Assisted Stress Corrosion Cracking

• Irradiated Concrete and Containment

• Electrical Cable Qualification and Condition Assessment 11

RPV Neutron Embrittlement at High Fluence

• Neutron fluence of PBAPS Units 2 and 3 at 80 years estimated at 2.23 x 1018 n/cm2

- Not considered high fluence

• Reactor Vessel Material Surveillance Program (B.2.1.20) enhancement

- Unit 2: withdraw 120 degree capsule (reconstituted specimens) at 60-62 EFPY

- Unit 3: withdraw 120 degree capsule at 60-62 EFPY

• Evaluation of time-limited aging analyses in SLRA Section 4.2

- Implements Regulatory Guide 1.99, Revision 2, including consideration of surveillance program data 12

Reactor Internals Irradiation-Assisted Stress Corrosion Cracking (IASCC)

• IASCC of reactor vessel internals managed by:

- BWR Vessel Internals program (B.2.1.7)

- VT-1 and EVT-1 examinations to detect cracks

- Water Chemistry program (B.2.1.2)

• SLRA Section 4.2.14 Core Shroud IASCC and Embrittlement Analysis

- License renewal application (for 60 years) Section 4.3.2.2 identified a threshold value of 5.0E+20 n/cm2 (E > 1 MeV) for IASCC and embrittlement

- Fluence for core shroud and top guide projected to exceed this value for 80 years

- Dispositioned in accordance with 10 CFR 54.21(c)(1)(iii)

- Inspections of core shroud (BWRVIP-76-R1-A) and top guide (BWRVIP-26-A) 13

Irradiation of Concrete and Steel 14

Irradiation of Concrete and Steel Concrete

• The shield wall concrete provides radiation shielding and only the lower 10 ft of the wall concrete performs a structural function

• SRP-SLR thresholds are not exceeded at the peak location

• A plant-specific program is not needed Steel

• Used NUREG-1509 transition temperature approach

• 80-yr nil-ductility transition temperature (NDTT) is less than the lowest service temperature (LST) of 100° F

• Loss of fracture toughness is not an aging effect that needs to be managed for the RPV structural steel component supports 15

Electrical Cable Qualification and Condition Assessment

• Peach Bottom has installed the Smartcover' system on electrical cable manholes to monitor water accumulation:

- Float-less transmitter design for level monitoring and alarm

- Self-monitoring capability

- Alarms are generated to alert operators if high water level is sensed or a malfunction occurs

- Staff accepted five-year inspection frequency as an exception for annual inspection per GALL-SLR

• EQ qualified cables are covered under the EQ program per 10 CFR 50.49

- Exelon used the provisions of re-analysis per 10 CFR 50.49 to monitor and extend the qualified life of the cables

- The staff audited the re-analysis methods and calculations and confirmed consistency with current rules and guidelines 16

SLRA Review Conclusion

• On the basis of its review of the SLRA and the resolution of the confirmatory item, the staff determined that the requirements of 10 CFR 54.29(a) have been met for the subsequent license renewal of Peach Bottom Atomic Power Station Units 2 and 3.

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Region I AMP Inspections License Renewal Inspection Program for Period of Extended Operations Inspection Dates Results U2 IP 71003 4Q 2012 No Findings Phase 1 ML13029A013 U3 IP 71003 Not performed N/A Phase 1 U2 IP 71003 March 2013 No Findings Phase 2 ML13071A608 U3 IP71003 4Q 2014 No Findings Modified Phase 2 ML14121A474 U2 and U3 IP71003 4Q 2018 No Findings Phase 4 ML18355A401 18

Region I AMP Inspections AMPs Reviewed During 71004 Phase 4 Inspection

• Flow Accelerated Corrosion Program (existing)

• Maintenance Rule Structural Monitoring Program (existing)

• Ventilation System Inspection and Testing Activities (enhanced)

• Outdoor, Buried and Submerged Component Inspection Activities (enhanced)

• Fire Protection Activities (enhanced)

• In-accessible Medium Voltage Cables not subject to 10 CFR 50.49 Environmental Qualification Requirements (New) 19

Region I AMP Inspections ROP Baseline Inspections Inspection Date Age Management Program IP71111.08 ISI Annually alternate units Reactor Pressure Vessel and Internals ISI Program IP71111.07T Heat Sink 2013, 2016, 2019 GL 89-13 Activities Heat Exchanger Inspection Activities IP71111.21M DBAI 4Q 2017 Ensure the selected SSCs that are subject (operating in the post-40-year licensing period) to aging management review pursuant to 10 CFR Part 54 are being managed for aging in accordance with appropriate aging management programs.

IP71152 PI&R Sample 1Q 2018 The structural monitoring program did not Maintenance Rule Structural Monitoring Program match the procedural requirements for all elements IP71152 PI&R Sample 4Q 2015 Review of Unit 2 Torus Coating Defects Torus Water Chemistry IP71152 PI&R Sample 2Q 2018 Inaccessible Medium-Voltage Cables not subject to 10 CFR Cable reliability program 50.49 Environmental Qualification Requirements IP71152 PI&R Sample 4Q 2018 External flood seal Maintenance Rule Structural Monitoring Program IP71152 PI&R Sample 2Q 2019 U-3, Static O-Ring Pressure Switch Failure 10 CFR 50.49 Environmental Qualification Requirements 20

Region I AMP Inspections Resident Inspector Insight and Inspection Results

• No findings from License Renewal Program inspections and the AMPs being appropriately implemented

• Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, involving Exelons E-1 EDG underground cable failure

• Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving High Pressure Coolant Injection (HPCI) System Exhaust Pressure Switches Exceeded Documented Qualified Life 21

Region Conclusion

• Regional Inspections The inspectors found the aging management programs were being implemented in accordance with the license condition. The region will continue to monitor AMPs using the baseline Reactor Oversight Process.

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Summary Conclusion

• Completes the staffs presentation on the safety review of the Peach Bottom SLRA as documented in the Safety Evaluation Report with Confirmatory Item and the Region I presentation on inspections and plant material conditions

• Additional questions 23