ML19182A112

First 10 CFR 54.21 (B) Annual Amendment to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application

ML19182A112
Person / Time
Site:	Peach Bottom
Issue date:	07/01/2019
From:	Gallagher M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML19182A112 (24)

Text

Exelon Generation . 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50 10 CFR 51 10 CFR 54 July 1, 2019 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

First 10 CFR 54.21 (b) Annual Amendment to the Peach Bottom Atomic Power Station, Units 2 and 3, Subsequent License Renewal Application

Reference:

Letter from Michael P. Gallagher, Exelon Generation Company LLC (Exelon) to NRC Document Control Desk, dated July 1O, 2018, "Application for Subsequent Renewed Operating Licenses" In the Reference letter, Exelon Generation Company, LLC (Exelon) submitted the Subsequent License Renewal Application (SLRA) for the Peach Bottom Atomic Power Station (PBAPS),

Units 2 and 3.

Exelon has completed a review to identify any current licensing basis (CLB) changes made since submittal of the Reference letter, which have a material effect on the content of the SLRA, including the FSAR Supplement. This amendment identifies four (4) changes to the CLB that are considered to materially affect the contents of the PBAPS SLRA.

The enclosure to this letter contains a description of these CLB changes, and the corresponding mark-ups to the portions of the SLRA affected by the changes, thereby supplementing the PBAPS SLRA.

This submittal satisfies the 10 CFR 54.21 (b) requirement to submit an annual amendment to the SLRA for 2019, as well as the requirement to submit an amendment addressing any such changes at least three months before scheduled completion of the NRC review of the SLRA.

There are no new or revised regulatory commitments contained in this letter.

If you have any questions, please contact Mr. David J. Distel, Licensing Lead, PBAPS Subsequent License Renewal Project, at 610-765-5517.

July 1, 2019 U.S. Nuclear Regulatory Commission Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 1st day of July 2019.

Respectfully submitted, Vice President - License Renewal and Decommissioning Exelon Generation Company, LLC

Enclosure:

Current Licensing Basis Changes that Impact the SLRA Associated with the First 10 CFR 54.21 (b) Annual Amendment cc: Regional Administrator - NRC Region I NRC Project Manager (Safety Review), NRR-DMLR NRC Project Manager (Environmental Review), NRR-DMLR NRC Project Manager, NRR-DORL - Peach Bottom Atomic Power Station NRC Senior Resident Inspector, Peach Bottom Atomic Power Station R.R. Janati, Pennsylvania Bureau of Radiation Protection D.A. Tancabel, State of Maryland

July 1, 2019 Enclosure Page 1 of 22 Enclosure Peach Bottom Atomic Power Station, Units 2 and 3 Current Licensing Basis Changes that Impact the SLRA Associated with the First 10 CFR 54.21(b) Annual Amendment Introduction This Enclosure contains the descriptions of the four changes that are being made to the Subsequent License Renewal Application (SLRA) as a result of the 10 CFR 54.21 (b) annual amendment incorporating Current Licensing Basis (CLB) changes that materially affect the contents of the PBAPS SLRA. For each item, the change is described and the affected page number(s) and portion(s) of the SLRA is provided. For clarity, entire sentences or paragraphs from the SLRA are provided with deleted text highlighted by strikethroughs and inserted text highlighted by balded italics. Revisions to SLRA tables are shown by providing excerpts from the affected tables.

July 1, 2019 Enclosure Page 2 of 22 Change # 1 - Reactor Water Cleanup CRWCU) System Sample Chiller Relocation and Chiller Condenser Replacement Affected SLRA Sections: Table 2.3.3-21, Table 2.3.3-25, Section 3.3.2.1.21, Section 3.3.2.1.25, Table 3.3.1, Table 3.3.2-21, Table 3.3.2-25 SLRA Page Numbers: 2.3-96, 2.3-106, 3.3-28, 3.3-33, 3.3-141, 3.3-148, 3.3-317, 3.3-340 Description of Change:

Both PBAPS Unit 2 and Unit 3 RWCU System sample chillers were relocated within the Reactor Buildings. As a result, the liquid filled sample chiller process pumps and associated piping are no longer in an enclosure and are added to the scope of the Process Sampling System for spatial interaction with a leakage boundary intended function. In addition to relocating the chillers, the air cooled condensers were replaced with water cooled condensers. The water heat sink for the condensers is provided by the Reactor Building Closed Cooling Water (RBCCW) System. The liquid filled water cooled condensers, which are also not contained within an enclosure, are added to the scope of the RBCCW System for spatial interaction with a leakage boundary intended function.

All other liquid filled components associated with the sample chillers remain within an enclosure and therefore are not in scope for spatial interaction.

Accordingly, SLRA Table 2.3.3-21, Table 2.3.3-25, Section 3.3.2.1.21, Section 3.3.2.1.25, Table 3.3.1, Table 3.3.2-21, and Table 3.3.2-25 are revised.

July 1, 2019 Enclosure Page 3 of 22 SLRA Table 2.3.3-21, Process Sampling System Components Subject to Aging Management Review on page 2.3-96 is revised as shown below:

Table 2.3.3-21 Process Sampling System Components Subject to Aging Management Review

July 1, 2019 Enclosure Page 4 of 22 SLRA Table 2.3.3-25, Reactor Building Closed Cooling Water System Components Subject to Aging Management Review on page 2.3-106 is revised as shown below:

Table 2.3.3-25 Reactor Building Closed Cooling Water System Components Subject to Aging Management Review Component Tvpe Intended Function Bolting (Closure) Mechanical Closure Compressor Housing (Instrument Leakage Boundary Nitrogen Compressor)

Heat Exchanger - (Instrument Nitrogen Leakage Boundary Compressor Aftercooler) Tube Side Components Heat Exchanger - (Instrument Nitrogen Leakage Boundary Compressor Aftercooler) Tubes Heat Exchanger - (PASS Jet Pump and Leakage Boundary Liquid Sample Coolers) Shell Side Components Heat Exchanger - (RBCCW Heat Leakage Boundary Exchangers) Shell Side Components Heat Exchanger - (RWCU Non- Leakage Boundary Regenerative Heat Exchanger) Shell Side Components Heat Exchanger - (RWCU Pump Motor Leakage Boundary Cooler) Shell Side Comoonents Heat Exchanger - (RWCU Sample Leakage Boundary Chiller Condenser) Tubes Heat Exchanger - (Reactor Water Sample Leakage Boundary Heat Transfer Coil) Shell Side Components Hoses Leakage Boundary Insulated Valve Body Leakage Boundary Insulated piping, piping components Leakage Boundary Piping elements Leakage Boundary Piping, piping components Leakage Boundary Pump Casing (RBCCW Pumps) Leakage Boundary Tanks (RBCCW Chemical Addition Tank) Leakage Boundary Tanks (RBCCW Head Tank) Leakage Boundary Valve Body Leakage Boundary

July 1, 2019 Enclosure Page 5 of 22 SLRA Section 3.3.2.1.21, Process Sampling System on page 3.3-28 is revised as shown below:

3.3.2.1.21 Process Sampling System Environments The Process Sampling System components are exposed to the following environments:

• Air - Indoor Uncontrolled

• Closed Cycle Cooling Water

• Condensation

• RawWater

• Treated Water

• Treated Water> 140 F

• Waste Water

July 1, 2019 Enclosure Page 6 of 22 SLRA Section 3.3.2.1.25, Reactor Building Closed Cooling Water System on page 3.3-33 is revised as shown below:

3.3.2.1.25 Reactor Building Closed Cooling Water System Environments The Reactor Building Closed Cooling Water System components are exposed to the following environments:

• Air - Indoor Uncontrolled

• Closed Cycle Cooling Water

• Condensation

• Gas

July 1, 2019 Enclosure Page 7 of 22 SLRA Table 3.3.1, Summary of Aging Management Evaluations for the Auxiliary Systems, Items 3.3.1-205 and 3.3.1-232 on pages 3.3-141 and 3.3-148 is revised as shown below:

Table 3.3.1 Summary of Aging Management Evaluations for the Auxiliary Systems Item Component Aging Effect/ Aging Management Further Discussion Number Mechanism Programs Evaluation Recommended 3.3.1-205 Insulated stainless steel Cracking due to SCC AMP Xl.M29, "Outdoor Yes Consistent with NUREG-2191. The One-piping, piping and Large Atmospheric Time Inspection (B.2.1.21) program will be components, tanks Metallic Storage Tanks," used to manage cracking of insulated exposed to air, AMP Xl.M32, "One-Time stainless steel piping, piping components condensation Inspection," AMP Xl.M36, exposed to air-outdoor and condensation in "External Surfaces the Chilled Water System, Emergency Monitoring of Mechanical Service Water System, Process Sampling Components," or AMP System, Refueling Water Storage and Xl.M42, "Internal Transfer System, and Service Water Coatings/Linings for In- System.

Scope Piping, Piping See Subsection 3.3.2.2.3.

Components, Heat Exchangers, and Tanks" 3.3.1-232 Insulated stainless steel, Loss of material due AMP Xl.M29, "Outdoor Yes Consistent with NUREG-2191 . The One-nickel alloy piping, piping to pitting, crevice and Large Atmospheric Time Inspection (B.2.1.21) program will be components, tanks corrosion Metallic Storage Tanks," used to manage loss of material of the exposed to air, AMP Xl.M32, "One-Time insulated stainless steel piping, piping condensation Inspection," AMP Xl.M36, components exposed to air-outdoor and "External Surfaces condensation in the Chilled Water System, Monitoring of Mechanical Emergency Service Water System, Components," or AMP Process Sampling System, Refueling Xl.M42, "Internal Water Storage and Transfer System, and Coatings/Linings tor In- Service Water System.

Scope Piping, Piping See Subsection 3.3.2.2.4.

Components, Heat Exchangers, and Tanks"

July 1, 2019 Enclosure Page 8 of 22 SLRA Table 3.3.2-21, Process Sampling System, Summary of Aging Management Evaluation on page 3.3-317 is revised as shown below:

Table 3.3.2-21 Process Sampling System Component Intended Material Environment Aging Effect Aging Management NUREG-2191 NUREG-2192 Notes Type Function Requiring Programs Item Table 1 Item Management Bolting (Closure) Mechanical Closure Carbon and Low Air- Indoor Loss of Material Bolting Integrity (B.2.1.10) Vll.l.A-03 3.3.1-012 B Alloy Steel Uncontrolled (External)

Bolting Loss of Preload Bolting Integrity (B.2.1.10) Vll.l.AP-124 3.3.1-015 B Insulated piping, Leakage Boundary Copper Alloy Condensation None None Vl/.J.AP-144 3.3.1-114 A piping with 15% Zinc 01 (External) components Less Treated Water Loss of Material One-Time Inspection Vll.E3.AP-140 3.3.1-022 A (Internal) (B.2.1.21)

Water Chemistry Vll.E3.AP-140 3.3.1-022 B (B.2.1.2)

Insulated Pump Leakage Boundary Stainless Steel Condensation Cracking One-Time Inspection Vll.l.A-734b 3.3.1-205 A Casing (RWCU (External) (B.2.1.21)

Sample Chiller)

Loss of Material One-Time Inspection Vll.l.A-761b 3.3.1-232 A (B.2.1.21)

Treated Water Loss of Material One-Time Inspection Vll.E4.AP-110 3.3.1-203 A

{Internal) (B.2.1.21)

Water Chemistry Vll.E4.AP-110 3.3.1-203 B (B.2.1.2)

July 1, 2019 Enclosure Page 9 of 22 SLRA Table 3.3.2-25, Reactor Building Closed Cooling Water System, Summary of Aging Management Evaluation on page 3.3-340 is revised as shown below:

Table 3.3.2-25 Reactor Building Closed Cooling Water System (Continued)

Component Intended Material Environment Aging Effect Aging Management NUREG-2191 NUREG-2192 Notes Type Function Requiring Programs Item Table 1 Item ManaQement Heat Exchanger - Leakage Boundary Carbon Steel Closed Cycle Cooling Loss of Material Closed Treated Water Vll.C2.AP-189 3.3.1-046 B (RWCU Pump Water (Internal) Systems (B.2.1.12)

Motor Cooler) Shell Side Comoonents Heat Exchanger- Leakage Boundary Copper Alloy Closed Cycle Cooling Loss of Material Closed Treated Water Vll.C2.AP-199 3.3.1-046 D (RWCU Sample with 15% Zinc oi Water (Internal) Systems (B.2.1.12)

Chiller Less Condenser) Gas (External) None None Vll.J.AP-9 3.3.1-114 c Tubes Heat Exchanger - Leakage Boundary Copper Alloy with Air - Indoor None None VIJ.J.AP-144 3.3.1-114 A (Reactor Water 15% Zinc or Less Uncontrolled (External)

Sample Heat Closed Cycle Cooling Loss of Material Closed Treated Water Vll.C2.AP-199 3.3.1-046 B Transfer Coil) Shell Water (Internal) Systems (B.2.1.12)

Side Components

July 1, 2019 Enclosure Page 10 of 22 Change# 2- Fuel Pool Cooling and Cleanup System Heat Exchanger Replacement Affected SLRA Sections: Tables 3.3.1, 3.3.2-16 SLRA Page Numbers: 3.3-96, 3.3-279, 3.3-280 Description of Change:

The Fuel Pool Cooling heat exchangers have been replaced, and the heat exchanger shells for the new heat exchangers are a different material than the original heat exchangers. SLRA Table 3.3.2-16, "Fuel Pool Cooling and Cleanup System," identifies the original fuel pool cooling heat exchanger shells as carbon steel; however the material for the replacement heat exchanger shells is stainless steel.

Accordingly, SLRA Tables 3.3.1 and 3.3.2-16 are updated to reflect the new material.

July 1, 2019 Enclosure Page 11 of 22 SLRA Table 3.3.1, Summary of Aging Management Evaluations tor the Auxiliary Systems, page 3.3-96, is revised as shown below:

Table 3.3.1 Summary of Aging Management Evaluations for the Auxiliary Systems Item Component Aging Effect/ Aging Management Further Discussion Number Mechanism Programs Evaluation Recommended 3.3.1-080 Steel heat exchanger Loss of material due AMP Xl.M36, "External No Consistent with NUREG-2191. The components, piping, to general, pitting, Surfaces Monitoring of External Surfaces Monitoring of Mechanical piping components crevice corrosion Mechanical Components" Components (B.2.1 .24) program will be exposed to air - indoor used to manage loss of material of the uncontrolled, air - carbon steel, ductile iron, and gray cast iron outdoor heat exchanger components, piping, piping components exposed to air-indoor uncontrolled and air-outdoor in the Auxiliary Steam System, Chilled Water System, Emergency Cooling Water System, Emergency Diesel Generator System, Emergency Service Water System, ~

Pool CooliR!J amj CleaRl:li;J Syslem, High Pressure Service Water System, Reactor Building Closed Cooling Water System, Reactor Water Cleanup System, Service Water System, and Turbine Building Closed Cooling Water System.

July 1, 2019 Enclosure Page 12 of 22 SLRA Table 3.3.2-16, Fuel Pool Cooling and Cleanup System, Summary of Aging Management Evaluation, pages 3.3-279 and 3.3-280, is revised as shown below:

Table 3.3.2-16 Fuel Pool Cooling and Cleanup System Component Intended Material Environment Aging Effect Aging Management NUREG-2191 NUREG-2192 Notes Type Function Requiring Programs Item Table 1 Item Manaaement Heat Exchanger - Leakage Boundary GaF89A Steel Air - Indoor Cracking One-Time Inspection Vl/.E4.AP-209a 3.3.1-004 c (Fuel Pool Cooling Uncontrolled (External) (B.2.1.21)

Heat Exchanger)

Shell Side Stainless Steel Components Loss of Material One-Time Inspection Vll.E4.AP-221 a 3.3.1-006 c (B.2.1.21)

E>GeFAal S1:1rfases VII.I.Ar;> 41 3.3.1 oso A MeAiteFiAg et Mesl=laAisal r /f"'J "' of ..... ~\

• - *- ,-*-* *-

Treated Water (Internal) beAg +eFm bess et GAe +ime IASf:!estieA VI 1..1\4 ..A. 43Q 3.3.1 1ga A IC ., < <H \

Mate Fial ,-*-* * - I Loss of Material One-Time Inspection Vll.A4.AP-111 3.3.1-203 A (B.2.1.21) VI I. E4 .*l\12 100 3.3.1 Q21 G Water Chemistry (B.2.1.2) Vll.A4.AP* 111 3.3.1-203 B Vll.E4 .Al2 106 3.3.1 021 Q

July 1, 2019 Enclosure Page 13 of 22 Change # 3 - Feedwater System High/Low Pressure Interface Affected SLRA Sections: 2.3.4.3, Table 3.4.2-3 SLRA Page Numbers: 2.3-137, 3.4-77 Description of Change:

An additional fire protection function has been identified for several valves in the feedwater system. These valves perform a high/low pressure interface function for Appendix R. These valves are already in the scope of second license renewal and are subject to aging management as identified in SLRA Table 3.4.2-3, stainless steel valve bodies with leakage boundary function exposed to Air - Indoor Uncontrolled (External) and Treated Water> 140 F (Internal) environments. SLRA Section 2.3.4.3 and Table 3.4.2-3 are revised to include the pressure boundary function to reflect this fire protection function.

Accordingly, SLRA Section 2.3.4.3 and Table 3.4.2-3 are revised.

July 1, 2019 Enclosure Page 14 of 22 SLRA Section 2.3.4.3, Feedwater System, page 2.3-137, is revised as shown below:

2.3.4.3 Feedwater System Intended Functions

6. Relied upon in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for Fire Protection (1 O CFR 50.48). The Feedwater System provides an injection path into the reactor pressure vessel for both HPCI and RCIC. The Feedwater System also includes valves that are high/low pressure interfaces which are credited for Fire Safe Shutdown.

10 CFR 54.4(a)(3)

July 1, 2019 Enclosure Page 15 of 22 SLRA Table 3.4.2-3, Feedwater System, Summary of Aging Management Evaluation, page 3.4-77, is revised as shown below:

Table 3.4.2-3 Feedwater System (Continued)

Component Intended Material Environment Aging Effect Aging Management NUREG-2191 NUREG-2192 Notes Type Function Requiring Programs Item Table 1 Item Management Valve Body Pressure Boundary Stainless Steel Air - Indoor Cracking One-Time Inspection Vlll.D2.SP-118a 3.4.1-002 A Uncontrolled (External) (B.2.1.21)

Loss of Material One-Time Inspection Vlll.D2.SP-127a 3.4.1-003 A (B.2.1.21)

Treated Water (Internal) Loss of Material One-Time Inspection Vlll.02.SP-87 3.4.1-085 A (B.2.1.21)

Water Chemistry (B.2.1.2) Vlll.02.SP-87 3.4.1-085 B Treated Water> 140 f Cracking One-Time Inspection Vlll.E.SP-88 3.4.1-011 A

{Internal) (8.2.1.21J Water Chemistry Vlll.E.SP-88 3.4.1-011 B (B.2.1.2J Loss of Material One-Time Inspection Vl/l.D2.SP-87 3.4.1-085 A (B.2.1.21J Water Chemistry Vll/.D2.SP-87 3.4.1-085 B (B.2.1.2J

July 1, 2019 Enclosure Page 16 of 22 Change# 4-Transition to the Fifth Ten-Vear Interval of the ISi Program Affected SLRA Sections: A.2.1.5, B.2.1.1, B.2.1.3, B.2.1.5, B.2.1.30 SLRA Page Numbers: A-12, B-15, B-25, B-29, B-30, B-35, B-37, B-171 Description of Change:

The SLRA contains several references to the Fourth Ten-Year Interval of the ISi Program, which was the relevant Interval at the time that the SLRA was submitted. PBAPS is now in the Fifth Ten-Year Interval of the ISi Program. The SLRA is revised to reflect administrative changes that are relevant to the new Interval.

Accordingly, SLRA Sections A.2.1.5, B.2.1.1, B.2.1.3, B.2.1.5, and B.2.1.30 are revised.

July 1, 2019 Enclosure Page 17 of 22 SLRA Section A.2.1.5, page A-12, second paragraph, is revised as shown below:

A.2.1.5 BWR Stress Corrosion Cracking The program includes periodic volumetric examinations to detect and manage IGSCC in accordance with NRC GL 88-01. The extent and schedule of inspection described in GL 88-01 are modified in accordance with the inspection guidance in staff-approved BWRVIP-75-A.

Welds classified as IGSCC Category A may be inspected at a frequency in accordance with ASME Section XI, including the Code Case N-716-1 Risk Informed Inspection (RI /SI) program. subsumed into the Risk lnfoFmed lnseFVice Inspection pFogFam in accoFdance with sta.f:f apprnved EPRI Topical ~apart TR 112957, Revision B A, pending appFO'Jal of ASME Gode relief requests duFing the second peFiod of extended operation. The program includes the staff-approved positions delineated in NUREG-0313, Revision 2, and GL 88-01 and its Supplement 1 regarding selection of IGSCC resistant materials, solution heat treatment and stress improvement processes, water chemistry, weld overlay reinforcement, partial replacement, clamping devices, crack characterization and repair criteria, inspection methods and personnel, inspection schedules, sample expansion, leakage detection, and reporting requirements.

July 1, 2019 Enclosure Page 18 of 22 SLRA Section 8.2.1.1, page 8-15, fourth paragraph, is revised as shown below:

B.2.1.1 ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD For the current~ fifth 10-year inspection interval, the ISi program applies the requirements of ASME Code,Section XI, 200+ 2013 Edition. Edition throblgh 2003 Addenda, and Risk Informed lnservice Inspection (RI ISi) alternative reqblirements to Examination Categories Bi;, BJ, Ci; 1, and CF 2 for Class 1 and Class 2 piping 1Nelds as approved by relief reqblest. Examination locations, and the nblmber of locations reqbliring examination, are based on the gblidelines provided in EPRI TR 112857, "Revised Risk Informed lnservice Inspection Evalblation Procedblre," Revision B A, and ASME Code Case N 578 1. Examination Categories 8-F, B-J, C-F-1, C-F-2, C-A, and C-B piping welds have been exempted from ASME Section XI required surface and/or volumetric inspections by N-716-1. This Code Case allows for the implementation of the RI-IS/ Program.

July 1, 2019 Enclosure Page 19 of 22 SLRA Section B.2.1.3 is revised in several locations as shown below:

B.2.1.3 Reactor Head Closure Stud Bolting Page B-25, second paragraph The Reactor Head Closure Stud Bolting program implements ASME Code,Section XI inspection requirements through the ISi Program plan. The current ISi Program plan for the~ fifth 10-year inspection interval (November 5, 2008 January 2019 through December 31, 2018 2028) is based on the 200+ 2013 Edition of the ASME Code,Section XI, including 2003 addenda. The future 120-month inspection intervals will incorporate the requirements specified in the version of the ASME Code referenced in 10 CFR 50.55a 12 months before the start of the inspection interval.

Page B-29, second paragraph During this effectiveness review, examination reports for Unit 3 RPV head stud bushings 1-46 in 2011, and examination reports for flange threads 47-92 in 2015, could not be located. These examinations were recorded as complete in the ISi Program database. This issue was entered into the corrective action program. Because the examination reports could not be located, the inspection of RPV head stud bushings 1-46 was rescheduled and performed during the 2017 refueling outage, thereby ensuring that this required examination of the fourth 10-year ISi interval, covering the period from 2008 to 2018, was met. The inspection of flange threads 47-92 was not performed due to NRC approval of a relief request that has since eliminated the need to perform these inspections for the remainder of the current fourth 10-year ISi interval

(

Reference:

NRC Relief Request Approval, ADAMS Accession No. ML17170A013). The ISi Program database software that is currently in use addresses the cause of this issue and precludes recurrence by requiring all the necessary records associated with the completed examination to be entered into the database before the database statuses the examination as complete. An extent of condition review was also performed that determined there are no other similar issues.

Page B-30, fourth paragraph During the Unit 3 refueling outage in 2015, reactor head closure studs 47 through 92 were examined using the UT examination method. Reactor head closure nuts, washers, and bushings 47 through 92 were examined using the VT-1 method. There were no recordable indications. The Unit 3 RPV head stud bushings 1-46 were examined using the VT-1 method during the refueling outage in 2017. A Relief Request to ASME Code,Section XI requirements was approved in June 2017 that eliminated the need to perform UT examination of the flange threads for the remainder of the current fourth 10-year ISi interval.

July 1, 2019 Enclosure Page 20 of 22 SLRA Section B.2.1.5, page B-35, third paragraph, is revised as shown below:

B.2.1.5 BWR Stress Corrosion Cracking The program addresses the management of crack initiation and growth due to IGSCC in the piping, welds, and components through the implementation of the ISi program in accordance with ASME Code,Section XI. lnservice inspections, performed as augmented requirements of the Section XI ISi program, ensure that aging effects are identified and repaired before the loss of intended function of in scope components. The inspection frequency for welds classified in accordance with NRC GL 88-01 as IGSCC Category B through G is per the recommendations provided in the staff-approved BWRVIP-75-A, "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules" for normal water chemistry conditions. Welds classified as IGSCC Category A may be subsumed into tho Risk Informed lnsorvico Inspection (RI ISi) program in accordance with staff approved EPRI Topical Report TR 112057, Revision BA, Final Report, "Revised Risk Informed lnsorvico Inspection Evaluation Procedure," December 1QQQ, ponding appro\1od ASME Code relief request. In tho event that such relief is not approved by NRG staff for future ISi intervals during tho second period of extended operation, Category A 'Nolds 'Nould be examined per the extent and schedule defined by 8)PJRVIP 75 ft.inspected at a frequency in accordance with ASME Section XI, including tho Code Caso N 71 a 1 Risk Informed lnspestion (RI ISi) program. inspected at a frequency in accordance with ASME Section XI, including the Code Case N-716-1 Risk Informed Inspection (RI IS/) program.

July 1, 2019 Enclosure Page 21 of 22 SLRA Section B.2.1.5 is also revised on page 8-37 as shown below:

2. On Unit 2, during the fourth 10-year ISi inspection interval (November 2008 through December 2018), volumetric examinations are being performed on all 15 IGSCC Category D welds every six years and an examination was performed on one of the two IGSCC Category E welds. There are no IGSCC Category B, C, F, or G welds on Unit 2. Also, during this sblrront the fourth interval, 16 IGSCC Category A welds were examined. No indications of cracking were identified. Examinations of the Category D and E welds were performed per the schedules within BWRVIP-75-A. Examinations of the Category A welds were performed per the Risk-Informed ISi program schedules.

This example demonstrates that the industry guidelines delineated in NRC GL 88-01, NUREG-0313, Revision 2, and BWRVIP-75-A continue to be effectively implemented to monitor the condition of welds within the scope of the program.

3. On Unit 3, during the fourth 10-year ISi inspection interval, (November 2008 through December 2018), volumetric examinations were performed on two of the five IGSCC Category C welds and an examination was performed on the one IGSCC Category E weld. There are no IGSCC Category B, D, F, or G welds on Unit 3. Also, during this Sblrront the fourth interval, 22 IGSCC Category A welds were examined. No indications of cracking were identified.

Examinations of the Category C and E welds were performed per the schedules within BWRVIP-75-A. Examinations of the Category A welds were performed per the Risk-Informed ISi program schedules.

July 1, 2019 Enclosure Page 22 of 22 SLRA Section B.2.1.30, page B-171, sixth paragraph, is revised as shown below:

B.2.1.30 ASME Section XI, Subsection IWE The current program complies with ASME Section XI, Subsection IWE, 200+ 2013 Edition through the 2003 Addenda, supplemented with the applicable requirements of 10 CFR 50.55a.

In accordance with 10 CFR 50.55a(g)(4), the ISi program is updated each successive 120-month inspection interval to comply with the requirements of the latest edition of the ASME Code specified 12 months before the start of the inspection interval. The ASME Code edition consistent with the provisions of 10 CFR 50.55a will be used during the second period of extended operation.