DCL-15-150, Submittal of 10 CFR 54.21(b) Annual Update to the License Renewal Application (Lra), Amendment 51

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Submittal of 10 CFR 54.21(b) Annual Update to the License Renewal Application (Lra), Amendment 51
ML16004A149
Person / Time
Site: Diablo Canyon  
Issue date: 12/21/2015
From: Strickland L
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16004A158 List:
References
DCL-15-150, FOIA/PA-2016-0438
Download: ML16004A149 (44)
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{{#Wiki_filter:SPacific Gas and Electric Company L. Jearn Strickland, P.E. Diablo Canyon Power Plant Director P.O. Box 56 Technical Services Avila Beach, CA 93424 805.595.6476 El H8@pge.com December 21, 2015 PG&E Letter DCL-15-150 U.S. Nuclear Regulatory Commission 10 CFR 54.21(b) ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 10 CFR 54.21(b) Annual Update to the Diablo Canyon Power Plant License Renewal Application (LRA), Amendment 51

References:

1.

PG&E Letter DCL-09-079, "License Renewal Application," dated November 23, 2009

2.

PG&E Letter DCL-14-1 03, "10 CFR 54.21(b) Annual Update to the Diablo Canyon Power Plant License Renewal Application (LRA), Amendment 48 and LRA Appendix E, Applicant's Environmental Report - Operating License Renewal Stage, Amendment '1," dated December 22, 2014

3.

PG&E Letter DCL-15-121, "Response to NRC Letter dated September 24, 2015, Request for Additional Information for the Review of the Diablo Canyon Power Plant, Units 1 and 2, License Renewal Application - Set 38," dated October 21, 2015

Dear Commissioners and Staff:

By Reference 1, Pacific Gas and Electric Company (PG&E) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating Licenses DPR-80 and DPR-82, for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The application included the license renewal application (LRA) and LRA'Appendix E, "Applicant's Environmental Report - Operating License Renewal Stage." As required by 10 CFR 54.21(b), each year following submittal of the LRA, an update to the LRA must be submitted that identifies any change to the current licensing basis (CLB) that materially affects the contents of the LIRA, including the Final Safety Analysis Report Supplement. PG&E has not made a decision to move forward with the State licensing review process at this time. A schedule for potential coastal consistency review has not been established and will be provided if a decision is made to resume State licensing. A member of the STARS (Strategic Teaming and Resource Sharing) A[liance / Callaway e Diablo Canyon e Polo Verde

  • Wolf Creek

Document Control Desk PG&E Letter DCL-1 5-1 50 December 21, 2015 Page 2 identifies DCPP LRA changes that are being made to reflect CLB that materially affect the LRA. Enclosure 2 contains the affected LRA pages with changes shown as electronic markups (deletions crossed out and insertions italicized). The LRA update covers the period from October 1, 2014, through September 30, 2015. By Reference 2, PG&E committed to provide the NRC with responses to the applicable aging management program plant-specific action items, conditions, and limitations identified in the NRC Safety Evaluation (SE), Revision 1, on MRP-227 by December 2015. Enclosures 3 and 4 submit WCAP-1 7462-NP, Revision 1, "Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon Power Plant Unit 1,"and WCAP-17463-NP, Revision 1, "Program Plan for Aging Management of Reactor Vessel Internal at Diablo Canyon Power Plant Unit 2," respectively. WCAP-17462-NP, Revision 1, and WCAP-1 7463-NP, Revision 1, contain responses to the applicable plant-specific action items, conditions, and limitations identified in the NRC SE, Revision 1, on MRP-227. In addition, although not specifically required by the NRC SE, Revision 1, PG&E is aware that the NRC is requesting licensees to provide additional plant specific information to address NRC expectations and concerns regarding responses to plant-specific action items 1 and

2. Enclosure 5 provides additional plant-specific information that the NRC has been requesting from other licensees to address actions items I and 2 for DCPP Units 1 and 2. As noted in Enclosures 3 and 4, PG&E will be providing an evaluation of the DCPP Units 1 and 2 reactor internals components with regard to fuel designs and fuel management by March 31, 2016.

By Reference 3, PG&E committed to update the cathodic protection licensing basis by December 31, 2015. The design process for upgrading the cathodic protection system on buried, in-soil auxiliary saltwater system piping is ongoing. PG&E is revising this commitment to update the cathodic protection design and installation action plan and associated licensing basis by March 31, 2016. New and revised regulatory commitments (as defined by NEI 99-04) are provided in. Changes to existing LRA Table A4-1 commitments are contained in the changes to LIRA Table A4-1 in Enclosure 2. If you have any questions regarding this response, please contact Mr. Terence L. Grebel, License Renewal Project Manager, at (805) 458-0534. I have been delegated the authority of Edward D. Halpin, Senior Vice President - Power Generation and Chief Nuclear Officer, during his absence. I declare under penalty of perjury that the foregoing is true and correct. A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Document Control Desk December 21, 2015 Page 3 PG&E Letter DCL-1 5-1 50 Executed on December 21, 2015. Sincerely, / Director, Technical Services gwh/50668099 Enclosures cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator Siva P. Lingam, NRC Project Manager Richard A. Plasse, NRC Project Manager, License Renewal John P. Reynoso, Acting NRC Senior Resident Inspector Michael J. Wentzel, NRC Project Manager, License Renewal (Environ mental) A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Diablo Canyon
  • Palo Verde e Wolf Creek

Enclosure I PG&E Letter DCL-l15-1 50 Diablo Canyon Power Plant License Renewal Application (LRA) Changes Reflected in the Annual LRA Update Amendment 51 PG&E Letter DCL-1 5-1 50 Page 1 of 3 Diablo Canyon Power Plant License Renewal Application (LRA) Changes Reflected in the Annual LRA Update Amendment 51 Affected LRA Reason for Change Section Section 2.4.11 Updated to add a newly installed steel plate security-related enclosure to the scope of License Renewal. This enclosure was analyzed to ensure that failure of the enclosure will not impact Design Class I structures, systems, and components. Table 2.3.3-12 LIRA tables were updated to reflect the fire protection hose Table 3.3.2-12 stations as-built configuration. In the Diablo Canyon Power Plant (DCPP) LRA, hose stations represent the hose reel and not the associated isolation valve or piping; thus, there is no pressure boundary function. Hose reels only provide a structural support function. While verifying the hose reel material, it was noted that the copper alloy orifices and piping leading from the isolation valve to the fire hose were not addressed in the LRA. These components were added. Table 3.3.2-12 Errata. Updated to change aging effect of copper alloy (> 15% zinc) fire protection spray nozzles with an internal environment of plant indoor air from none to loss of material, consistent with NUREG-1 801, table item VII.G-9, as revised in LR-ISG-2012-02. Table 3.3.2-8 PG&E responded to license renewal request for additional information 2.3-2 related to whether the guard pipe enclosing the hydrogen piping was in the scope of license renewal in PG&E Letter DCL-10-067, "Response to License Renewal Application (LRA) Request for Additional Information and LRA Errata," dated June 18, 2010, and PG&E Letter DCL-10-128, "Response to NRC Letter dated September 13, 2010, Request for Additional Information (Set 23) for the Diablo Canyon License Renewal Application," dated October 12, 2010. Table 3.3.2-8 is updated to clarify that the guard pipe enclosing the hydrogen piping is not'a fire barrier since it does not prevent the spread of fire. The most-appropriate and most-conservative license renewal intended function of fire barrier was chosen to support its inclusion as being in-scope _________________because it is relied upon in the Fire Hazards Analysis.

Enclosure I PG&E Letter DCL-1 5-1 50 Page 2 of 3 Affected LRA Section Section 3.1.2.1.2 Section 3.3.2.1.3 Section 3.3.2.1.4 Section 3.3.2.1.5 Section 3.3.2.1.8 Section 3.3.2.1.12 Section 3.3.2.1.13 Section 3.3.2.1.17 Section 3.3.2.1.1 9 Section 3.4.2.1.1 Section 3.4.2.1.4 Reason for Change Errata. Updated to add the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks Program to manage the aging effects for the system component types. Section 3.2.2.1.1 Section 3.2.2.1.3 Section 3.3.2.1.3 Section 3.3.2.1.8 Section 3.4.2.1.3 Section 3.4.2.1.5 Errata. Updated to add wall thinning due to erosion as an aging effect requiring management, and where applicable, the Flow Accelerated Corrosion Program to manage the aging effects for the system component types. Section 3.2.2.1.2 Updated to reflect plant modifications (review of equipment Section 3.3.2.1.14 changes). Table 3.2.2-2 Table 3.3.2-3 Table 3.3.2-9 Table 3.3.2-14 Table 3.4.2-1 Section 4.7.5 Updated to reflect a new stress and fracture mechanics Section 4.9 evaluation completed to support request for alternative Section A3.5.3 REP-SI, Revision 2, "~Proposed Alternative to Requirements for Repair/Replacement Activities for Certain Safety Injection Pump Welded Attachments." Updated to reflect new inservice flaw growth analyses that were performed to address weld flaw indications identified in the Unit 2 structural weld overlays for the pressurizer safety spray nozzle welds. Updated to reflect a new inservice flaw growth analysis that was performed to address weld flaw indications identified in _________________the Unit I pressurizer spray line pipe weld. Section Al1.14 Updated to add fuel tank for portable diesel electric generators credited during fire protection events for _________________supporting safe shutdown. PG&E Letter DCLI15-150 Page 3 of 3 Affected LRA Reason for Change Section Section A1.13 In PG&E Letter DCL-14-1 03, "10 CFR 54.21 (b) Annual Table A4-1, Item 3 Update to the Diablo Canyon Power Plant License Renewal Application (LRA), Amendment 48 and LIRA Appendix E, Applicant's Environmental Report - Operating License Renewal Stage, Amendment 1," dated December 22, 2014, PG&E stated that following the enhancements listed in, Attachment 7C, and with the exceptions listed, the DCPP Fire Water System Program would be consistent with LR-ISG-2012-02, Section C. PG&E has determined that an enhancement not mentioned in DCL-14-103 was necessary for the deluge testing to be consistent with LR-ISG-2012-02, Section C. PG&E is revising the licensing basis to enhance the testing of deluge system nozzles consistent with the 2011 edition of NFPA 25, Section 10.3.4.3.1, as recommended by LR-ISG-2012-02, S Appendix D, Table 4a. Following this enhancement and those listed in DCL-14-1 03, Enclosure 1, Attachment 7C, and with the exceptions listed, PG&Es Fire Water System Program will be consistent with LR-ISG-2012-02, Section C. Table A4-1, Items 34 Updated the status of these items to show them as and 60 completed. Table A4-1, Item 73 of this letter provides DCPP Unit I and 2 responses to the applicable aging management program plant-specific action items, conditions, and limitations identified in the NRC SE, Revision 1, on MRP-227. Table A4-1, item 73 is revised to show this commitment is _________________ complete. PG&E Letter DCL-1 5-1 50 License Renewal Application (LRA) Amendment 51 Affected LRA Sections and Tables, and Figures PG&E Letter DCL-1 5-150 Page 1 of 29 License Renewal Application (LRA) Amendment 51 Affected LRA Sections and Tables, and Figures Table 2.3.3-12 Section 2.4.11 Section 3.1.2.1.2 Section 3.2.2.1.1 Section 3.2.2.1.2 Section 3.2.2.1.3 Table 3.2.2-2 Section 3.3.2.1.3 Section 3.3.2.1.4 Section 3.3.2.1.5 Section 3.3.2.1.8 Section 3.3.2.1.12 Section 3.3.2.1.13 Section 3.3.2.1.14 Section 3.3.2.1.17 Section 3.3.2.1.19 Table 3.3.2-3 Table 3.3.2-8 Table 3.3.2-9 Table 3.3.2-12 Table 3.3.2-14 Section 3.4.2.1.1 Section 3.4.2.1.3 Section 3.4.2.1.4 Section 3.4.2.1.5 Table 3.4.2-1 Section 4.7.5 Section 4.9 Section AI. 13 Section A1.14 Section A3.5.3 Table A4-1, Items 3, 34, 60, and 73 PG&E Letter DCL-1 5-1 50 Page 2 of 29 Table 2.3.3-12 Fire Protection System Section 2.3 SCOPING AND SCREENING RESULTS MECHANICAL SYSTEMS Component Type Intended Function Bellows Pressure Boundary Closure Bolting Pressure Boundary Flow Element Pressure Boundary Flow Indicator Pressure Boundary Hose Station P'ressure..... d.. 'Structural Support Hydrant Pressure Boundary Orifice Pressure Boundary __________________________Throttle Piping Leakage Boundary (spatial) Pressure Boundary __________________________Structural Support Pump Pressure Boundary __________________________Structural Support RCP Oil Collection Reservoir Pressure Boundary Solenoid Valve Pressure Boundary Spray Nozzle Spray Strainer Pressure Boundary Tank Pressure Boundary __________________________Structural Support Test Connection Pressure Boundary Trailer Structural Support Tubing Leakage Boundary (spatial) _________________________Pressure Boundary Valve Leakage Boundary (spatial) Pressure Boundary Vessel Pressure Boundary Section 2.4 PG&E Letter DCL-1 5-150 SCOPING AND SCREENING RESULTS Page 3 of 29 STRUCTURES 2.4.11 Earthwork and Yard Structures Structure Description The earthwork and yard structures include the circulating water conduits, auxiliary saltwater (ASW) vacuum breaker vaults, ASW thrust blocks and anchors, a security-re/a ted enclosure, raw water storage reservoirs 1A and 1 B, east and west breakwaters, and the earth slopes east of the auxiliary building and over the ASW line east of the intake structure. The seismically qualified portions of the circulating water conduits and ASW vacuum breaker vaults are reinforced concrete structures founded on compacted fill. The Design Class I ASW supply piping is supported by reinforced concrete thrust blocks, compacted backfill, and concrete anchors attached to the circulating water conduits. The seismically qualified security-related enclosure is a steel structure and was analyzed to ensure that failure of the enclosure will not impact Design Class I SSCs. The security-related enclosure does not perform any (a)(1) intended functions and does not contain components required by the five License Renewal regulated events (a) (3). The raw water reservoir, located east of the power block, has reinforced concrete-walls. The reservoir is primarily intended to serve as fresh water storage for fire protection and long term cooling. The breakwater structures, which are constructed of precast reinforced concrete blocks and rip-rap, protect the intake structure from tsunami loads. The earth slopes east of auxiliary building and over the ASW line east of the intake structure were analyzed for design basis seismic loads to ensure that such loading will not produce any significant slope failure that can impact Design Class I SSCs. The ASW system buried piping and electrical conduits are protected from tsunami/storm conditions by wave protection measures, which include concrete covers, revetments, roadway slabs, and pavement. Gabion mattresses embedded~within the slopes are covered with grass for additional erosion control. For the purposes of license renewal and aging management, the breakwaters and earth slope protection structures are evaluated as barriers. Structure Intended Functions The earthwork and yard structures, except the security-related enclosure, provide structural support, shelter, and protection for components relied upon to provide the capability to shutdown the reactor and maintain it in a safe shutdown condition. The raw water reservoir provides fresh water storage for long term cooling. The earthwork and yard structures, except the security-related enclosure, also provide structural support, shelter, and protection for nonsafety-related SSCs whose Section 2.4 PG&E Letter DCL-1 5-1 50 SCOPING AND SCREENING RESULTS Page 4 of 29 STRUCTURES failure could prevent performance of a safety-related function. Therefore, the structures are within the scope of license renewal based on the criterion of 10 CFR 54.4(a)(2). The security-related enclosure provides structural support whose failure could prevent performance of a safety-related function. Therefore, the structure is within the scope of license renewal based on the criterion of 10 CFR 54.4(a)(2). The earthwork and yard structures, except the security-related enclosure, provide structural support, shelter, and protection for components required to support fire protection and SBO requirements. Therefore, the earthwork and yard structures are within the scope of license renewal based on the criteria of 10 CFR 54.4(a)(3). Section 3.1 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF REACTOR VESSEL Page 5 of 29 INTERNALS, AND REACTOR COOLANT SYSTEM 3.1.2.1.2 Reactor Coolant System Aging Management Programs The following aging management programs manage the aging effects for the reactor coolant system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) Section 3.2 PG&E Letter DCL-1 5-1 50 AGING MANAGEMENT OF ENGINEERED Page 6 of 29 SAFETY FEATURES 3.2.2.1.1 Safety Injection System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the saltwater and chlorination system component types: Flow-Accelerated Corrosion (B2. 1.6) 3.2.2.1.2 Containment Spray System Materials The materials of construction for the containment spray system component types are: 3.2.2.1.3 Residual Heat Removal System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the saltwater and chlorination system component types: Flow-Accelerated Corrosion (B2.1.6) PG&E Letter DCL-15-150 Page 7 of 29 Section 3.2 AGING MANAGEMENT OF ENGINEERED SAFETY FEATURES Table 3. 2.2-2 Engineered Safety Features - Summary of Aging M anagement Evaluation - Containment Spra y System Component Intended Material Environment Aging Effect Aging Management NUREG-Table I Item Notes Type Function Requiring Program 1801 Vol. Management 2 Item VaIve R-B G*,n"f-A (!nt) No None V--F--4 3*~4-5 A Valve R-B Gop-lo PatIdo Ai Non Ne* V-F-A Section 3.3 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 8 of 29 3.3.2.1.3 Saltwater and Chlorination System Materials The materials of construction for the saltwater and chlorination system component types are: Ductile Iron Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the saltwater and chlorination system component types: Flow-Accelerated Corrosion (B2. 1.6) Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.3.2.1.4 Component Cooling Water System Aging Management Programs The following aging management programs manage the aging effects for the component cooling water system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2.1. 42) 3.3.2.1.5 Makeup Water System Aging Management Programs The following aging management programs manage the aging effects for the makeup water system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (82.1.42) Section 3.3 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 9 of 29 3.3.2.1.8 Chemical and Volume Control System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the chemical and volume control system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1. 42) 3.3.2.1.12 Fire Protection System Aging Management Programs The following aging management programs manage the aging effects for the fire protection system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.3.2.1.13 Diesel Generator Fuel Oil System Aging Management Programs The following aging management programs manage the aging effects for the diesel generator fuel oil system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components. Heat Exchangers, and Tanks (B2. 1.42) 3.3.2.1.14 Diesel Generator System Materials The materials of construction for the diesel generator system component types are: Copper Alloy (> 15 percent Zinc) Section 3.3 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 10 of 29 3.3.2.1.17 Liquid Radwaste System Aging Management Programs The following aging management programs manage the aging effects for the liquid radwaste system component types: S Internal Coatings/Linings for in-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.3.2.1.19 Oily Water and Turbine Sump System Aging Management Programs The following aging management programs manage the aging effects for the oily water and turbine sump system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 142) PG&E Letter DCL-1 5-1 50 Page 11 oft29 SECTION 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3. 3.2-3 Auxiliary Systems - Summary of Aging Management Evaluation - Saltwater and Chlorination System Component Intended Material Environment Aging Effect Aging Management NUREG-Table 1 Item Notes Type Function Requiring Program 1801 Vol. _______Management 2 Item Separator LBS Nickel Alloys Plant Indoor Air None None VII.J-14 3.3.1.94 A ~(Ext) Separator LBS Nickel Alloys Raw Water (lnt) Loss of material Open-Cycle Cooling VII. C1-13 3.3.1.78 A Water System (B2. 1.9) Valve LBS Ductile Iron Plant Indoor Air Loss of material External Surfaces VIII-8 3.3.1.58 B (Ext) Monitoring Program (B2. 1.20) Valve LBS Ductile Iron Raw Water (Int) Loss of material Open-Cycle Cooling VII. C1-19 3.3.1.76 A Water System (B2. 1.9) Valve LBS Ductile Iron Raw Water (Int) Wall thinning due Flow-Accelerated None None H, 5 ____to erosion Corrosion (B2. 1.6)_ PG&E Letter DCL-15-150 Page 12 of 29 SECTION 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS T~hlk R.* 2-R A~iviliart, 5Zv~efmjc -- 5Zlimmnrv nf Atninn AAn~ninmimnt ~I::inhIi~tin -- flht.mir'l ~nnr I l/AIm* Crinfrrl.VIQfctim Component Intended Material Environment Aging Effect Aging Management Program NUREG-Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item Piping FB, LBS, Carbon Plant Indoor Air Loss of material External Surfaces Monitoring VII.I-8 3.3.1.58 8, 12 PB, SIA Steel Ex)Program (B2.1.20) Piping FB, LBS, Carbon Plant Indoor Air Loss of material Inspection of Internal Surfaces V.A-19 3.2.1.32 B3, 12 SIA Steel (Int) in Miscellaneous Piping and Ducting Components (B2.1.22) Notes for Table 3.3.2-8: Plant Specific Notes: 12 The guard pipe enclosing the hydrogen piping is credited in the Fire Hazards Analysis and is thus in the scope of License Renewal per 10 CFR 54. 4(a)(3), The guard pipe is not a fire barrier since it does not prevent the spread of fire, as defined in Regulatory Guide 1.120. The most-appropriate and most-conservative license renewal intended function of "fire barrier" was chosen to support its inclusion as being in-scope because it is relied upon in the Fire Hazards Analysis. Table 3.3.2-9 Auxiliary Systems - Summary of Aging Management Evaluation - Miscellaneous HVAC Systems Component Intended Material Environment Aging Effect Aging Management NUREG-Table I Item Notes Type Function Requiring Program 1801 Vol. Management 2 Item Valve SIA, SS Copper Alloy Plant Indoor Air None None VIII.I-2 3.4.1.41 C ________Ext) Valve SIA, SS Copper Alloy Ventilation Loss of material Inspection of Internal VII.G-9 3.3.1.28 iE Atmosphere (Int) Surfaces in Miscellaneous Piping and Ducting _________Components_(B2.1.22) PG&E Letter DCL-15-150 Page 13 of 29 SECTION 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-12 Auxiliary Systems - Summary of Aging Management Evaluation - Fire Protection System Component Intended Material Environment Aging Effect Aging Management Program NUREG-Table I Notes Type Function Requiring 1801 Vol. Item Management 2 Item Hose Station P-BSS Carbon Steel Atmosphere/ Loss of material External Surfaces Monitoring VII.I-9 3.3.1.58 B, 16 Weather (Ext) Program (B2. 1.20) Hose Station P-BSS Carbon Steel Plant Indoor Air Loss of material External Surfaces Monitoring VII.I-8 3.3.1.58 B, 16 (Ext) Program (B2.1.20) Hose-Station P-B Gabn-te R ae-{* Loss ofmateria! Fire Water Sys*tem, (B2.1-1-3 V-G). 34-448S B Hose-Station P-B Cabn-te Ra-aer-Tt ruroenl Fire Water System (B2.!.13-) None None Gorrosio

OiiePB, TH Copper Alloy Atmosphere!

Loss of material Selective Leaching of None None G Oiie(> 15% Zinc) Weather (Ext Materials (B2. 1.17)

OiiePB, TH Copper Alloy Plant Indoor Air None None VIII.I-2 3.4.1.41 A

Oriice(> 15% Zinc) Ext

OiiePB, TH Copper Alloy Plant Indoor Air Loss of material Fire Water System (B2. 1.13)

VlI.G-9 3.3.1.28 E, 8 Oriice(> 15% Zinc) (Int Piping PB Copper Alloy Atmosphere! Loss of material Selective Leaching of None None G "> 15% Zinc) Weather (Ext Materials (B2. 1.17) Piping PB Copper Alloy Plant Indoor Air None None VIII.I-.2 3.4.1.41 A '> 15% Zinc) ESxt) PB Copper Alloy Plant Indoor Air Loss of material Fire Water System (B2. 1.13) VII.G-9 3.3.1.28 E, 8 Piping (> 15% Zinc) (Int) Spray Nozzle SP Copper Alloy Plant Indoor Air Loss of Fire Water System VII. G-3.3.1.28Nort E, 8G (_______> 15% Zinc) (Int) materialNepne (B2. 1. 13)Nene 9Nene-e SECTION 3.3 PG&E Letter DCL-15-150 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 14 of 29 Notes for Table 3.3.2-12: Plant Specific Notes: 8 The Fire Water System program (B2.1 13) is used to monitor copper alloy piping, piping components and piping elements exposed to condensation (internal) for loss of material in the fire protection system. Reference LR-ISG-2012-02, Appendix C, Line VII.G.A-1 43, aind PG&E Letter DCL-14-1 03, Enclosure 1, Attachment 7C, and DCL-15-150, Enclosure 2.

16.

This line item only represents the hose reel and not the associated isolation valve, piping, or fittings. PG&E Letter DCL-15-150 Page 15 of 29 SECTION 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-14 Auxiliary Systems - Summary of Aging Management Evaluation - Diesel Generator System Component Intended Material Environment Aging Effect Aging Management NUREG-Table I Item Notes Type Function Requiring Program 1801 Vol. ______Management 2 Item_____ Valve PB Aluminum Plant Indoor Air Loss of material Inspection of Internal VII.F2-12 3.3. 1.27 E (Int) Surfaces in Miscellaneous Piping and Ducting Components (B2. 1.22) Valve PB Aluminum Plant Indoor Air None None VII. J-1 3.3.1.95 A (Ext) Valve PB Copper Alloy Lubricating Oil Loss of material Lubricating Oil Analysis VII. H2-10 3.3.1.26 B (> 15% Zinc) (Int) (B2. 1.23) and One-Time Inspection (B2. 1.16) Valve PB Copper Alloy Plant Indoor Air None None V.F-3 3.2.1.53 A ______(> 15% Zinc) (Ext)_______ Section 3.4 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF STEAM AND Page 16 of 29 POWER CONVERSION SYSTEM 3.4.2.1.1 Turbine Steam Supply System Aging Management Programs The following aging management programs manage the aging effects for the turbine steam supply system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.4.2.1.3 Feedwater System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion 3.4.2.1.4 Condensate System Aging Management Programs The following aging management programs manage the aging effects for the condensate system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.4.2.1.5 Auxiliary Feedwater System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the saltwater and chlorination system component types: S Flow-Accelerated Corrosion (B2. 1.6) PG&E Letter DCL-1 5-1 50 Page 17 of 29 SECTION 3.4 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Table 3.4.2-1 Steam and Power Conversion System - Summar'y of Aging Management Evaluation - Turbine Steam Sunoltv System Component Intended Material Environment Aging Effect Aging Management NUREG-Table I Item Notes Type Function Requiring Program 1801 Vol. Management 2 Item Expansion LBS Nickel Alloys Plant Indoor Air None None VIII.I-9 3.4.1.41 A Joint (Ext) Expansion LBS Nickel Alloys Steam (Int) Loss of material Water Chemistry VIII. B1-1 3.4.1.37 E, 4 Joint (B2. 1.2) and One-Time ____ Inspection (B2. 1.16) Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 18 of 29 4.7.5 Inservice Flaw Growth Analyses that Demonstrate Structural Stability for 40 Years Summary Description The 1SI procedure states that a fracture mechanics analysis, in accordance with ASME Section XI Code, Subsection IWB-3600, must be completed if a flaw acceptance criterion is not met as outlined in the corresponding test procedure. These analyses depend on a specified number of operating years, and thus may be TLAAs for DCPP. Analysis Unit 2 RHR Piping Weld RB-119.11 During a routine inservice inspection prior to DCPP Unit 2 Refueling Outage 13 (2R1 3) in 2006, a circumferential flaw was identified in Weld RB-I119-11 of the residual heat removal (RHR) system. The observed flaw did not meet the Section Xl acceptance standards of Table IWB-3514-2. Consequently, the indication was evaluated per the guidelines of Section Xl, IWB-3640. A conservative fatigue crack growth evaluation was then performed to determine the adequacy of the piping system for continued operation. The evaluation was submitted to the NRC for review, as required by the Code, in PG&E Letter DCL-06-069. The service life for Weld RB-i119-11 is based on operating for 40 years from the date the flaw was identified, i.e. until 2046, during which the flaw would experience 500 startup-shutdown cycles. Thus, the evaluation encompassed a 60-year plant life and the analysis will be valid beyond the 2045 end date of the period of extended operation for Unit 2. The cycle assumptions used in the analysis are conservative compared to the DCPP original design cycles described in Section 4.3.1.1. The DCPP licensing basis assumes 250 heatups and 250 cooldowns for a 50 year plant life. Since the analysis indicates that the allowable flaw depth will not be reached for the next 40 years of plant operation beginning in October 2006, the flaw evaluation of RHR Weld RB-i119-11 will remain valid for the period of extended operation in accordance with 10 CFR 54.211l(1)(i). Unit 2 Auxiliary Feedwater Piping Line 567 During Unit 2 Refueling Outage 8 (2R8), while performing a non-routine surface examination prior to maintenance, DCPP identified a flaw indication in the auxiliary feedwater pump recirculation header Line 567, that exceeds Section Xl, Table IWB-3410-1 criteria. The flaw has been accepted by analysis by meeting the Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 19 of 29 allowable size criteria of IWB-3620 and IWB-3610 and was submitted to the NRC in PG&E Letter DCL-99-1 36. The numbers of thermal and seismic cycles used in the analysis are consistent with or more conservative than the DCPP 50-year design basis described in FSAR Table 5.2-4. The assumed transients are consistent with or bounded by the 50 year licensing basis. The number of transients will be monitored by the enhanced Fatigue Management Program. The enhanced Fatigue Management Program provides assurance that the fatigue crack growth analysis will be managed for the period of extended operation in accordance with 10 CFR 54.211(1 )(iii). Unit I RHR Piping Weld WIC-95 During Unit 1 Refueling Outage 9 (1R9), while performing an inservice inspection, DCPP identified a weld flaw indication located in an ASME Class 2 portion of the RHR injection Line 985 to hot legs 1 and 2 at weld WLC-95. The indication exceeded the Section Xl, Table IWC-3410-1 criteria. The flaw has been accepted by analysis in accordance with IWB-3410 and was submitted to the NRC in PG&E Letter DCL-97-086. The number of seismic cycles used in the analysis is consistent with the DCPP 50-year design basis described in FSAR Table 5.2-4. There have been no occurrences of a DE, DDE, or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the Unit 1 RHR Weld WIC-95 fatigue crack growth evaluation for the 50-year design basis number of DE, DDE, and Hosgri events is sufficient to the end of the period of extended operation. Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CFR 54.211l(1)(i). Units I and 2 Safety Injection Pumps Vent and Drain Socket Welds In 2014, PG&E requested NRC approval of lnservice Inspection Request for Alternative REP-SI for DCPP, Units 1 and 2. To support this request, a stress and fracture mechanics evaluation was performed to determine the adequacy of the socket welds associated with ASME Class 2 Safety Injection (SI) Pumps 1-1, 1-2, and 2-1 vent and drain connections. The evaluation was submitted to the NRC for review, as part of the relief request, in PG&E Letter DCL-14-060, dated July 21, 2014. Postulated flaws were evaluated using a fracture mechanics approach analogous to the methods of ASME Code Section XI. This relief request was approved for the remaining life of the subject SI Pumps, including the duration of the current operating licenses plus a license renewal period of 20 years [Reference 42]. Since the evaluation is based on the 60-year operating period, the TLAA covers the period of extended operation and is dispositioned under 10 CFR 54.2 11(1)(i). Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 20 of 29 For the postulated crack analysis, 7, 000 thermal transient cycles (pump starts), 400 DE cycles (20 events with 20 cycles per event), and 20 Hosgri earthquake cycles (1 event with 20 cycles) were assumed. Using a conservative projection of 1,400 SI Pump start cycles for a 60 year plant life, the 7, 000 thermal transient cycles assumed in the postulated crack analysis during 60 years of operation is conservative. The number of seismic cycles used in the analysis is consistent with the DCPP 50-year design basis described in FSAR Table 5.2-4. There have been no occurrences of a DE or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the postulated crack analysis for the 50-year design basis number of DE and Hosgri events is sufficient to the end of the period of extended operation. The analysis is valid for the period of extended operation in accordance with 10 CFR 54.211(1)(i). Unit 2 Pressurizer Safety and Spray Nozzle Welds As stated in LRA Section 4. 7.2, during Unit 2 Refueling Outage 14 (2R 14, Spring 2008), Alloy 690 structural weld overlays were completed on Alloy 82/182 welds attaching the surge, spray, and relief valve nozzles to the safe ends, and the safe ends to the connecting piping. During the seventeenth Unit 2 Refueling Outage (2R1 7), while performing inservice inspections, DCPP identified weld flaw indications located at Unit 2structural weld overlays for the pressurizer safety nozzles Aand B, and pressurizer spray nozzle. Conservative fatigue crack growth evaluations were then performed to determine the adequacy of the piping system for continued operation. The evaluations were submitted to the NRC for review, as part of a relief request, in PG&E Letter DCL 028, dated April 7, 2014. This relief request was approved for the service life of the structural weld overlays [Reference 43]. The service life for the pressurizer safety and spray nozzle structural weld overlays is based on operating for 38 years from the date the structural weld overlays were completed, i e. until 2046. Thus, the evaluation encompassed a 60-year plant life and the analysis will be valid beyond the 2045 end date of the period of extended operation for Unit 2. The cycle assumptions used in the analyses are consistent with those transients used in the pressurizer structural weld overlay (LRA Section 4.7.2 and PG&E Letter DCL-10-120). Per LRA Table A 4-1, Commitment 38, the plant transient cycles related to the structural weld overlay fatigue crack growth analyses are included in the existing plant transient monitoring program. Since the analyses indicate that the allowable flaw depth will not be reached for the remaining plant life, the flaw evaluations of the pressurizer safety and spray nozzle structural weld overlays will remain valid for the period of extended operation in accordance with 10 CFR 54.211(1)(i). Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 21 of 29 Unit I Pressurizer Spray Line Pipe Weld WIB-378 During Unit 1 Refueling Outage 19 (1R 19), while performing an inservice inspection, DCPP identified a weld flaw indication located in an ASME Code Class 1 pressurizer spray line pipe weld WlB-3 78. The indication exceeded the Section Xl, Table IWB-35 14-2 criteria. The flaw has been accepted by analysis in accordance with iWB-3600 and was submitted to the NRC in PG&E Letter DCL-15-131, dated November 3, 2015. A fatigue crack growth evaluation was performed to determine the adequacy of the weld for continued operation. The indication was a planar flaw oriented circumferentially and was assumed to be ID connected for conservatism. The service life for Weld WIB-378 is based on operating through the period of extended operation. Thus, the evaluation encompassed a 60-year plant life and the analysis will remain valid for the Unit 1 period of extended operation. The number of transient cycles used in the analysis is consistent with or more conservative than the DCPP 50-year design basis described in FSAR Table 5.2-4. Because the evaluation indicates that the allowable flaw depth will not be reached for the remaining plant life and the assumed transients are consistent with or more conservative than the DCPP 50-year design basis described in FSAR Table 5. 2-4, the flaw evaluation of the pressurizer spray line pipe weld WlB-378 will remain valid for the period of extended operation in accordance with 10 CFFR 54.2 11(1)(i). Disposition: Validation, 10 CFR 54.21 1(1 )(i); and Aging Management, 10 CFR 54.211l(1)(iii) Validation - Flaw Evaluation of Unit 2 RHR Piping Weld RB-119-II The result indicates that the allowable flaw depth will not be reached for the next 40 years of plant operation beginning in October 2006. Therefore, the flaw evaluation of RHR Weld RB-I119-11 will remain valid for the period of extended operation in accordance with 10 CFR 54.21 I(1)(i). Validation - Flaw Evaluation of Unit I RHR Weld WIC-95 There have been no occurrences of a DE, DDE, or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the Unit 1 RHR Weld WIC-95 fatigue crack growth evaluation for the 50-year design basis number of DE, DDE, and Hosgri events is sufficient to the end of the period of extended operation. Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CER 54.211(1)(i). Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 22 of 29 Validation - Units 1 and 2 Safety Injection Pumps Vent and Drain Socket Welds Using a conservative projection of 1,400 SI Pump start cycles for a 60 year plant life, the 7, 000 thermal transient cycles assumed in the postulated crack analysis during 60 years of operation is conservative. There have been no occurrences of a DE or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the postulated crack analysis for the 50-year design basis number of DE and Hosgri events is sufficient to the end of the period of extended operation. The analysis is valid for the period of extended operation in accordance with 10 CFR~ 54.211(1)(i). Validation - Flaw Evaluation of Unit 2 Pressurizer Safety and Spray Nozzle Welds The results indicate that the allowable flaw depth will not be reached for the remaining plant life. Therefore, the flaw evaluation of Unit 2 Pressurizer Safety and Spray Nozzle Welds will remain valid for the period of extended operation in accordance with 10 CFR 54.2 11(1)(i). Validation - Flaw Evaluation of Unit I Pressurizer Spray Line Pipe Weld WIB-378 The results indicate that the allowable flaw depth will not be reached for the remaining plant life. Therefore, the flaw evaluation of Unit 1 Pressurizer Spray Line Pipe Weld W/B-378 will remain valid for the period of extended operation in accordance with 10 CFR 54.211(1)(i). Aging Management - Unit 2 Auxiliary Feedwater Piping Line 567 The Metal Fatigue of the Reactor Coolant Pressure Boundary program (B3. 1) monitors fatigue design transients including the transients assumed in the fatigue crack growth analyses for the Unit 2 auxiliary feedwater piping Line 567. The program provides assurance that the fatigue crack growth analysis will be managed for the period of extended operation in accordance with 10 CFR 54.211(1 )(iii). Section 4 PG&E Letter DCL-15-150 TIME-LIMITED AGING ANALYSES Page 23 of 29

4.9 REFERENCES

42.

US NRC Letter. From Michael T. Markley, Chief, Plant Licensing Branch IV-l, Division of Operating Reactor Licensing, Off~ice of Nuclear Reactor Regulation; to Mr. Edward D. Halpin, Senior Vice President and Chief Nuclear Officer, DCPP. "Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Request for Alternative REP-SI, Revision 2, Proposed Alternative to Requirements for Repair/Replacement Activities for Certain Safety Injection Pump Welded Attachments (TAC Nos. MF4476 and MF44 77)." 15 July 2015. (ML15187A035)

43.

US NRC Letter. From Michael T. Markley, Chief, Plant Licensing Branch IV-1, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation; to Mr. Edward D. Halpin, Senior Vice President and Chief Nuclear Officer, DCPP. "Diablo Canyon Power Plant, Unit No. 2 - Inservice Inspection Program Relief Request SWOL-REP-1 U2 for Approval of an Alternative to the ASME Code, Section XI, for Preemptive Full Structural Weld Overlays (TAC No. MF389 1)." 14 October 2014. (ML14255A232) Appendix A PG&E Letter DCL-1 5-150 Final Safety Analysis Report Supplement Page 24 of 29 AI. 13 FIRE WATER SYSTEM The Fire Water System program manages loss of material due to corrosion, including MIC, fouling, flow blockage because of fouling, and loss of integrity for water-based fire protection systems and internal coatings/linings for the fire water storage tank within the scope of license renewal. Internal and external inspections and tests of fire protection equipment are performed consistent, with exceptions identified in PG&E Letters DCL-14-1 03, Enclosure 1, Attachment 7C, and DCL-15-121 with NFPA-25 (2011 edition). Testing or replacement of sprinklers that have been in place for 50 years is performed in accordance with NFPA-25 (2011 edition). Portions of the deluge systems that are normally dry but periodically subjected to flow and cannot be drained or allow water to collect will undergo augmented testing beyond that in NFPA-25 consisting of volumetric wall thickness examinations. The fire water system is managed by performing routine preventive maintenance, inspections and testing; operator rounds, performance monitoring, and reliance on the corrective action program; and system improvements to address aging and obsolescence issues. The fire water system is normally maintained at required operating pressure and is monitored such that loss of system pressure is immediately detected and corrective actions are initiated. The Fire Water System program will conduct a flow test with air, water, or other medium through each open spray nozzle to verify that deluge systems nozzles are unobstructed. Water flow tests will verify that the deluge system provide full coverage of the equipment it protects. Visual inspections will be performed on firewater piping. Non-intrusive follow-up volumetric examinations will be performed if internal visual inspections detect surface irregularities to determine if wall thickness is within acceptable limits. Visual inspections will evaluate for the presence of sufficient foreign material to obstruct fire water pipe or sprinklers. Inspections of the firewater tank will be performed to detect loss of material. As discussed in PG&E Letter DCL-1 5-027, Enclosure 1, in response to LR-ISG-2013-01, the program consists of periodic visual inspections of the internal liner of the fire water storage tank exposed to raw water where loss of lining integrity could impact the components' and downstream components' current licensing basis intended function(s). For coated surfaces determined to not meet the acceptance criteria, physical testing is performed where physically possible (i.e., sufficient room to conduct testing) in conjunction with repair, replacement, or removal of the lining. The training and qualification of individuals involved in coating inspections are conducted in accordance with ASTM International Standards endorsed in RG 1.54 including guidance from the NRC associated with a particular standard. The Fire Water program implements the recommendations in LR-ISG-2012-02, as discussed in PG&E Letters DCL-14-1 03, Enclosure 1, Attachments 7C,-arn4 DCL-15-121, and DCL-15-150, the recommendations in LR-ISG-2013-01, as discussed in PG&E Letter DCL-1 5-027 Appendix A PG&E Letter DCL-1 5-150 Final Safety Analysis Report Supplement Page 25 of 29 A1.14 FUEL OIL CHEMISTRY The Fuel Oil Chemistry program manages loss of material on the internal surface of components in the emergency diesel fuel oil storage and transfer system, portable diesel electric generator fuel oil tanks, portable diesel driven fire pump fuel oil tanks, and portable caddy fuel oil tanks. The program includes (a) surveillance and monitoring procedures for maintaining fuel oil quality by controlling contaminants in accordance with applicable ASTM Standards, (b) periodic draining of water from fuel oil tanks, (c) visual inspection of internal surfaces during periodic draining and cleaning, (d) one-time ultrasonic wall thickness measurements of accessible portions of fuel oil tank bottoms, (e) sampling and analysis of new fuel oil before it is introduced into the fuel oil tanks, and (f) supplemental one-time inspections of a representative sample of components in systems that contain fuel oil by the One-Time Inspection program (A1.16). Appendix A PG&E Letter DCL-1 5-150 Final Safety Analysis Report Supplement Page 26 of 29 A3.5.3 Inservice Flaw Growth Analyses that Demonstrate Structural Stability for 40 Years The ISI procedure states that a fracture mechanics analysis, in accordance with ASME Code, Section Xl, Subsection IWB-3600, must be completed if flaw acceptance criterion is not met as outlined in the corresponding test procedure. These analyses depend on a specified number of operating years, and thus may be TLAAs. Unit 2 RHR Piping Weld RB-119-11 In 2006, a circumferential flaw was identified in DCPP Unit 2 Weld RB-i119-11 of the residual heat removal (RHR) system. The observed flaw did not meet the Section XI acceptance standards of Table IWB-3514-2. Consequently, the indication was evaluated per the guidelines of Section Xl, IWB-3640. A conservative fatigue crack growth evaluation was performed to determine the adequacy of continued operation of the piping system. The analysis is based on operating for 40 years from the date the flaw was identified and will be valid beyond the end of the period of extended operation for Unit 2 in accordance with 10 CFR 54.21 1(1 )(i). Unit 2 Auxiliary Feedwater Piping Line 567 DCPP identified a flaw indication in the Unit 2 auxiliary feedwater pumps recirculation header Line 567, that exceeds Section Xl, Table IWB-3410-1 criteria. The flaw has been accepted by analysis by meeting the allowable size criteria of IWB-3620 and IWB-3610. The Metal Fatigue of Reactor Coolant Pressure Boundary program described in Section A2.1I monitors fatigue design transients including the transients assumed in the fatigue crack growth analyses and therefore will be managed for the period of extended operation in accordance with 10 CFR 54.211l(1)(iii). Unit I RHR Piping Weld WIC-95 DCPP identified a weld flaw indication located in an ASME Class 2 portion of the Unit 1 residual heat removal injection Line 985 to hot legs 1 and 2 at weld WIC-95. The indication exceeded the Section Xl, Table IWC-341 0-1 criteria. The flaw has been accepted by analysis in accordance with IWB-3410. The number of seismic cycles assumed in the analysis is sufficient for the period of extended operation. Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CFR 54.211(1)(i). Appendix A PG&E Letter DCL-1 5-150 Final Safety Analysis Report Supplement Page 27 of 29 Units 1 and 2 Safety Injection Pumps Vent and Drain Socket Welds In support of a relief request, a stress and fracture mechanics evaluation was performed to determine the adequacy of socket welds associated with ASME Class 2 Safety Injection (SI) Pumps 1-1, 1-2, and 2-1 vent and drain connections. Postulated flaws were evaluated using a fracture mechanics approach analogous to the methods of ASME Code Section Xl. The number of cycles used in the analysis is sufficient for the period of extended operation. Using a conservative projection of 1,400 SI Pump start cycles for a 60 year plant life, the 7, 000 thermal transient cycles assumed in the postulated crack analysis during 60 years of operation is conservative. Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CFR 54.2 11(1)(i). Unit 2 Pressurizer Safety and Spray Nozzle Welds In 2013, laminar flaws were identified in DCPP Unit 2 structural weld overlays for pressurizer safety nozzles A and B, and pressurizer spray nozzle. Conservative fatigue crack growth evaluations were performed to determine the adequacy of continued operation of the piping system. The analyses are based on operating for 38 years from the date the structural weld overlays were completed and will be valid beyond the end of the period of extended operation for Unit 2 in accordance with 10 CFR 54.2 11(1)(i). Unit 1 Pressurizer Spray Line Pipe Weld WIB-378 In 2015, a circumferential flaw was identified in DCPP Unit 1 Pressurizer Spray Line Pipe Weld WIB-3 78. The observed flaw did not meet the Section Xl acceptance standards of Table IWB-3514-2. Consequently, the indication was evaluated per the guidelines of Section XI, IWB-3600. A conservative fatigue crack growth evaluation was performed to determine the adequacy of continued operation of the wveld. The analysis is based on operating through the period of extended operation and will remain valid Unit 1 in accordance with 10 CFR 54.211(1)(i). PG&E Letter DCL-15-150 Page 28 of 29 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Table A4-1I License Renewal Commitments LRA Implementation Item # Commitment Section Schedule 3 Enhance the Fire Water System program: B2.1.13 Program is (a) Sprinkler heads in service for 50 years will be replaced or representative samples from implemented 5 one or more sample areas will be tested consistent with NFPA 25, Inspection, Testing years before the and Maintenance of Water-Based Fire Protection Systems, 2011 Edition guidance. period of extended Test procedures will be repeated at 10-year intervals during the period of extended operation. operation, for sprinkler heads that were not replaced prior to being in service for 50 Inspections of years, to ensure that signs of degradation, such as corrosion, are detected prior to the wetted normally loss of intended function, and dry piping (b) To perform non-intrusive follow-up volumetric examinations if internal visual segments that inspections detect surface irregularities to determine if wall thickness is within cannot be drained acceptable limits. Visual inspections will evaluate for the presence of sufficient foreign or that allow water material to obstruct fire water pipe or sprinklers to collect begin 5 (c) To be in conformance with LR-ISG-2012-02, Section C as discussed in PG&E Letter years before the DCL-14-1 03, Enclosure 1, Attachment 7C. period of extended (d) To be in conformance with LR-ISG-2013-01 as discussed in PG&E Letter DCL-15-027, operation. Internal. linings inspections (e) Test deluge system nozzles in accordance with the 2011 Edition of NFPA 25, begin no later than Section 10.3.4.3. 1. the last refueling outage before the period of extended operation. The program's remaining inspections begin during the period of extended __________ operation 34 The DCPP work control procedure will be revised to include evaluation of reinforced concrete B1.232 Complete. PG&E exposed during excavations. Letter DCL-15-150 PG&E Letter DCL-1 5-1 50 Page 29 of 29 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Table A4-1 License Renewal Commitments LRA Implementation Item # Commitment Section Schedule 60 PG&E will enhance provisions in the HVAC ducting from the 480V switchgear room that allow water to drain from the exhaust ducting so water cannot enter the 480V switchgear room. Complete. PG&E Letter DCL 150..PfotG4 The NRC SE for MRP-227 contains eight action items for applicants/licensees to consider. Cmlt.P& Responses to the applicable aging management program plant-specific action items, Comp.41letter PG&E5-50 73 conditions, and limitations identified in the NRC SE, Revision 1, on MRP-227 will be submitted D214 ecemerDC2015-10 to the NRC by December 2015. Reference DCL-14-1 03, Enclosure 1, Attachment 4. PG&E Letter DCL-1 5-1 50 WCAP-1 7462-NP, Revision 1 Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon Power Plant Unit I PG&E Letter DCL-1 5-1 50 Page 1 of 5 MRP-227-A Applicability Guideline for Diablo Canyon Power Plant Westinghouse Pressurized Water Reactor Design

Background

The Nuclear Regulatory Commission (NRC) staff has determined that additional information, as discussed in References 1 and 2, should be provided by licensees to verify the applicability of MRP-227-A (Reference 6). The two specific generic issues that need to be addressed are summarized as follows:

1. Do the reactor internals have any non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater, and if so, do the affected components have operating stresses greater than 30 ksi?
2. Does the plant have atypical fuel design or fuel management that could render the assumptions of MRP-227-A, regarding core loading/core design, non-representative?

PG&E Response to Question I Diablo Canyon Power Plant (DCPP) Units 1 and 2 reactor internals components have been evaluated according to industry guideline MRP 2013-025 (Reference 3), as well as to the MRP-1 91 (Reference 4) industry generic component listings and screening criteria (including consideration of cold work as defined in MRP-175 (Reference 5), noting the requirements of Section 3.2.3). In addition to consideration of the material fabrication, forming, and finishing process, a general screening definition of "severe cold work" [a resulting reduction in wall thickness (material stock thickness) of 20 percent] was applied as an evaluation limit. The evaluation included a review of all plant modifications affecting reactor internals and the plant operating history. The components were procured according to American Society for Testing and Materials International of American Society of Mechanical Engineers material specifications that were callouts in the original plant construction drawings. Thus, material identification based on the material callouts and notes in the component drawings was an efficient and reasonable approach to identify the material of construction of components for DCPP Units 1 and 2. Based on the specifications used in the DCPP Units land 2 plant component drawings, it was possible to bin the reactor internals components into five material categories identified in MRP 2013-025. DCPP Units I and 2 components were binned according to the following categories for the materials used in the component fabrication. PG&E Letter DCL-1 5-1 50 Page 2 of 5 Categories based on MRP 2013-025 include:

  • Cast austenitic stainless steel (CASS) (Category 1)
  • Hot-formed austenitic stainless steel (Category 2)
  • Annealed austenitic stainless steel (Category 3)
  • Fasteners austenitic stainless steel (Category 4)
  • Cold-formed austenitic stainless steel without subsequent solution annealing (Category 5)

The potential for cold work is directly controlled by the materials specifications. Essentially, all of the components that are binned (based on their specified materials) as Categories 1, 2, and 3 are non-cold worked; therefore, they have less than 20 percent cold work according to NRC criterion. Similarly, any component binned under Category 5 has the potential to contain greater than 20 percent cold work. Category 4 materials are fasteners that may have been intentionally strain-hardened. The strain hardening according to guidelines should have been intentionally restricted to less than 20 percent. Material definitions in drawings identify maximum yield stress restrictions on these materials, which allows for the identification of the cold work level. In some cases, however, these restrictions are not present on drawings. Restrictions or limitations on the material yield stress (e.g., a maximum of 90 ksi) would indicate that the material cold work would be limited to less than 20 percent. In the absence of a maximum restriction yield stress of strain-hardened material, a conservative approach was taken to indicate the potential for greater than 20 percent cold work. Where multiple options existed for a component or assembly, the bounding condition was taken as the option that had the greater potential to include greater than 20 percent cold work. This. option was then employed in the assessment of the component and was selected for the purposes of the Westinghouse evaluation. In some instances, sequential fabrication would appear to mitigate any potential for cold work; however, since the historical record was not detailed, the potential is noted, but a conservative approach was selected for the Westinghouse evaluation. The evaluation, performed consistently with MRP 2013-025, concluded that the reactor internals Categories 1, 2, and 3 (non-bolting) components at DCPP Units I and 2 contain no cold work greater than 20 percent as a result of material specification and controlled fabrication construction. Category 4 components were already assumed to have the potential for coldwork in the MRP-191 generic assessments. No Category 5 components with severe cold work were identified for DCPP Units I and 2. The detailed evaluation for~the DCPP Units 1 and 2 cold work assessments concluded that the plant-specific fabrication and design was consistent with the PG&E Letter DCL-15-150 Page 3 of 5 MRP-1 91 basis, and that the MRP-227-A (Reference 6) sampling inspection aging management requirements, as related to cold work, are directly applicable to DCPP Units I and 2. The inspection sampling requirements for aging management outlined in MRP-232 (Reference 7) are based on the assumptions of MRP-191 and MRP-227-A. Therefore, MRP-232 calls for the demonstration that the plant specific materials, fabrication, and design meet the assumptions inherent in MRP-191 and MRP-227-A. The detailed Westinghouse evaluation of DCPP Units I and 2 material fabrication and design has concluded that no non-fastener materials of greater than 20 percent cold work were used in construction, and that the inspection sampling approach of MRP-232 is applicable to DCPP Units I and 2. PG&E Response to Question 2 As stated in MRP 2013-025, to demonstrate plant-specific applicability of the MRP-227-A sampling inspection strategy for managing aging in reactor internals, licen~sees must demonstrate that the criteria of MRP-227-A, Section 2.4 are met, and that the neutron fluence and heat generation rates are within the range of the following variables summarized. As detailed in WCAP-1 7462-NP, Revision 1, "Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon Power Plant Unit 1," and WCAP-17463-NP, Revision 1, "Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon Power Plant Unit 2," for DCPP Units I and 2, respectively, the criteria specified in MRP-227-A, Section 2.4 has been demonstrated as follows. The MRP-227-A, Section 2.4 assumptions are stated first, followed by' a description of how the assumptions are addressed at DCPP Units I and 2. The assumptions from MRP-227-A, Section 2.4 are as follows:

  • 30 years of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low leakage fuel management strategy for the remaining 30 years of operation; DCPP Units 1 and 2 fuel management programs changed from a high-to
  • low-leakage core loading pattern prior to 30 years of operation.
  • Base load operation, iLe., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule.

DCPP Units 1 and 2 operate as base-load units. PG&E Letter DCL-1 5-1 50 Page 4 of 5 No design changes beyond those identified in general industry guidance or recommended by the original vendors. MRP-227-A states that the recommendations are applicable to all U.S. PWR operating plants as of May 2007 for the three designs considered. There have been no modifications to reactor internals components at DCPP Units 1 or 2 since May 2007. Based on the applicability, as stated for DCPP Units 1 and 2, the criteria, of MRP-227-A, Section 2.4 are met for DCPP Units 1 and 2. In addition to the req*uirement to demonstrate that the criteria of MRP-227-A, Section 2.4 are met, MRP 2013-025 requires that the neutron fluence and heat generation rates for DCPP Units 1 and 2 are within the range of the limiting threshold values defined in MRP 2013-025. The limiting threshold values defined for Westinghouse plants are: Average core power density less than 124 Watts/cm3 Heat generation figure of merit (F) less than or equal to 68 Watts/cm3

  • Active fuel to upper core plate distance greater than 12.2 inches PG&E is currently in the process of evaluating the DCPP Units I and 2 reactor internals components with regard to fuel designs and fuel management according to guidance provided in MRP 2013-025. PG&E is currently scheduled to complete and submit to the NRC the results of this evaluation for DCPP Units 1 and 2 by March 31, 2016 (see Enclosure 4).

References

1. U.S. Nuclear Regulatory Commission Letter, "Summary of January 22-23, 2013, Closed Meeting with the Electric Power Research Institute and Westinghouse," February 21, 2013. (ADAMS: ML13042A048/mlI13043A062).
2. U.S. Nuclear Regulatory Commission Letter, "Summary of February 25, 2013, Telecom with the Electric Power Research Institute and Westinghouse Electric Company," March15, 2013. (ADAMS: ML13067A262).
3. EPRI Letter, MRP 2013-025, "MRP-227-A Applicability Template Guideline,"

October 14, 2013.

4. Materials Reliability Program: Screening, Categorization and Ranking of PWR Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

PG&E Letter DCL-1 5-1 50 Page 5 of 5

5. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175). EPRI, Palo Alto, CA: 2005. 1012081.
6. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA:

2011. 1022863.

7. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internal Components (MRP-232, Rev. 1). EPRI, Palo Alto, CA: 2012. 1021029.

PG&E Letter DCL-1 5-1 50 Regulatory Commitments Pacific Gas and Electric Company (PG&E) is making the following new and revised regulatory commitments (as defined by NEI 99-04) in this submittal: Commitment Due Date PG&E will update the cathodic protection design and installation action plan and associated licensing basis by March 31, 2016 March 31, 2016. PG&E is currently scheduled to complete and submit to March 31, 2016 the NRC an evaluation of the Units I and 2 reactor internals components with regard to fuel designs and fuel management ac~cording to guidance provided in MRP 2013-025.

SPacific Gas and Electric Company L. Jearn Strickland, P.E. Diablo Canyon Power Plant Director P.O. Box 56 Technical Services Avila Beach, CA 93424 805.595.6476 El H8@pge.com December 21, 2015 PG&E Letter DCL-15-150 U.S. Nuclear Regulatory Commission 10 CFR 54.21(b) ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 10 CFR 54.21(b) Annual Update to the Diablo Canyon Power Plant License Renewal Application (LRA), Amendment 51

References:

1.

PG&E Letter DCL-09-079, "License Renewal Application," dated November 23, 2009

2.

PG&E Letter DCL-14-1 03, "10 CFR 54.21(b) Annual Update to the Diablo Canyon Power Plant License Renewal Application (LRA), Amendment 48 and LRA Appendix E, Applicant's Environmental Report - Operating License Renewal Stage, Amendment '1," dated December 22, 2014

3.

PG&E Letter DCL-15-121, "Response to NRC Letter dated September 24, 2015, Request for Additional Information for the Review of the Diablo Canyon Power Plant, Units 1 and 2, License Renewal Application - Set 38," dated October 21, 2015

Dear Commissioners and Staff:

By Reference 1, Pacific Gas and Electric Company (PG&E) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating Licenses DPR-80 and DPR-82, for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The application included the license renewal application (LRA) and LRA'Appendix E, "Applicant's Environmental Report - Operating License Renewal Stage." As required by 10 CFR 54.21(b), each year following submittal of the LRA, an update to the LRA must be submitted that identifies any change to the current licensing basis (CLB) that materially affects the contents of the LIRA, including the Final Safety Analysis Report Supplement. PG&E has not made a decision to move forward with the State licensing review process at this time. A schedule for potential coastal consistency review has not been established and will be provided if a decision is made to resume State licensing. A member of the STARS (Strategic Teaming and Resource Sharing) A[liance / Callaway e Diablo Canyon e Polo Verde

  • Wolf Creek

Document Control Desk PG&E Letter DCL-1 5-1 50 December 21, 2015 Page 2 identifies DCPP LRA changes that are being made to reflect CLB that materially affect the LRA. Enclosure 2 contains the affected LRA pages with changes shown as electronic markups (deletions crossed out and insertions italicized). The LRA update covers the period from October 1, 2014, through September 30, 2015. By Reference 2, PG&E committed to provide the NRC with responses to the applicable aging management program plant-specific action items, conditions, and limitations identified in the NRC Safety Evaluation (SE), Revision 1, on MRP-227 by December 2015. Enclosures 3 and 4 submit WCAP-1 7462-NP, Revision 1, "Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon Power Plant Unit 1,"and WCAP-17463-NP, Revision 1, "Program Plan for Aging Management of Reactor Vessel Internal at Diablo Canyon Power Plant Unit 2," respectively. WCAP-17462-NP, Revision 1, and WCAP-1 7463-NP, Revision 1, contain responses to the applicable plant-specific action items, conditions, and limitations identified in the NRC SE, Revision 1, on MRP-227. In addition, although not specifically required by the NRC SE, Revision 1, PG&E is aware that the NRC is requesting licensees to provide additional plant specific information to address NRC expectations and concerns regarding responses to plant-specific action items 1 and

2. Enclosure 5 provides additional plant-specific information that the NRC has been requesting from other licensees to address actions items I and 2 for DCPP Units 1 and 2. As noted in Enclosures 3 and 4, PG&E will be providing an evaluation of the DCPP Units 1 and 2 reactor internals components with regard to fuel designs and fuel management by March 31, 2016.

By Reference 3, PG&E committed to update the cathodic protection licensing basis by December 31, 2015. The design process for upgrading the cathodic protection system on buried, in-soil auxiliary saltwater system piping is ongoing. PG&E is revising this commitment to update the cathodic protection design and installation action plan and associated licensing basis by March 31, 2016. New and revised regulatory commitments (as defined by NEI 99-04) are provided in. Changes to existing LRA Table A4-1 commitments are contained in the changes to LIRA Table A4-1 in Enclosure 2. If you have any questions regarding this response, please contact Mr. Terence L. Grebel, License Renewal Project Manager, at (805) 458-0534. I have been delegated the authority of Edward D. Halpin, Senior Vice President - Power Generation and Chief Nuclear Officer, during his absence. I declare under penalty of perjury that the foregoing is true and correct. A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Document Control Desk December 21, 2015 Page 3 PG&E Letter DCL-1 5-1 50 Executed on December 21, 2015. Sincerely, / Director, Technical Services gwh/50668099 Enclosures cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator Siva P. Lingam, NRC Project Manager Richard A. Plasse, NRC Project Manager, License Renewal John P. Reynoso, Acting NRC Senior Resident Inspector Michael J. Wentzel, NRC Project Manager, License Renewal (Environ mental) A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Diablo Canyon
  • Palo Verde e Wolf Creek

Enclosure I PG&E Letter DCL-l15-1 50 Diablo Canyon Power Plant License Renewal Application (LRA) Changes Reflected in the Annual LRA Update Amendment 51 PG&E Letter DCL-1 5-1 50 Page 1 of 3 Diablo Canyon Power Plant License Renewal Application (LRA) Changes Reflected in the Annual LRA Update Amendment 51 Affected LRA Reason for Change Section Section 2.4.11 Updated to add a newly installed steel plate security-related enclosure to the scope of License Renewal. This enclosure was analyzed to ensure that failure of the enclosure will not impact Design Class I structures, systems, and components. Table 2.3.3-12 LIRA tables were updated to reflect the fire protection hose Table 3.3.2-12 stations as-built configuration. In the Diablo Canyon Power Plant (DCPP) LRA, hose stations represent the hose reel and not the associated isolation valve or piping; thus, there is no pressure boundary function. Hose reels only provide a structural support function. While verifying the hose reel material, it was noted that the copper alloy orifices and piping leading from the isolation valve to the fire hose were not addressed in the LRA. These components were added. Table 3.3.2-12 Errata. Updated to change aging effect of copper alloy (> 15% zinc) fire protection spray nozzles with an internal environment of plant indoor air from none to loss of material, consistent with NUREG-1 801, table item VII.G-9, as revised in LR-ISG-2012-02. Table 3.3.2-8 PG&E responded to license renewal request for additional information 2.3-2 related to whether the guard pipe enclosing the hydrogen piping was in the scope of license renewal in PG&E Letter DCL-10-067, "Response to License Renewal Application (LRA) Request for Additional Information and LRA Errata," dated June 18, 2010, and PG&E Letter DCL-10-128, "Response to NRC Letter dated September 13, 2010, Request for Additional Information (Set 23) for the Diablo Canyon License Renewal Application," dated October 12, 2010. Table 3.3.2-8 is updated to clarify that the guard pipe enclosing the hydrogen piping is not'a fire barrier since it does not prevent the spread of fire. The most-appropriate and most-conservative license renewal intended function of fire barrier was chosen to support its inclusion as being in-scope _________________because it is relied upon in the Fire Hazards Analysis.

Enclosure I PG&E Letter DCL-1 5-1 50 Page 2 of 3 Affected LRA Section Section 3.1.2.1.2 Section 3.3.2.1.3 Section 3.3.2.1.4 Section 3.3.2.1.5 Section 3.3.2.1.8 Section 3.3.2.1.12 Section 3.3.2.1.13 Section 3.3.2.1.17 Section 3.3.2.1.1 9 Section 3.4.2.1.1 Section 3.4.2.1.4 Reason for Change Errata. Updated to add the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks Program to manage the aging effects for the system component types. Section 3.2.2.1.1 Section 3.2.2.1.3 Section 3.3.2.1.3 Section 3.3.2.1.8 Section 3.4.2.1.3 Section 3.4.2.1.5 Errata. Updated to add wall thinning due to erosion as an aging effect requiring management, and where applicable, the Flow Accelerated Corrosion Program to manage the aging effects for the system component types. Section 3.2.2.1.2 Updated to reflect plant modifications (review of equipment Section 3.3.2.1.14 changes). Table 3.2.2-2 Table 3.3.2-3 Table 3.3.2-9 Table 3.3.2-14 Table 3.4.2-1 Section 4.7.5 Updated to reflect a new stress and fracture mechanics Section 4.9 evaluation completed to support request for alternative Section A3.5.3 REP-SI, Revision 2, "~Proposed Alternative to Requirements for Repair/Replacement Activities for Certain Safety Injection Pump Welded Attachments." Updated to reflect new inservice flaw growth analyses that were performed to address weld flaw indications identified in the Unit 2 structural weld overlays for the pressurizer safety spray nozzle welds. Updated to reflect a new inservice flaw growth analysis that was performed to address weld flaw indications identified in _________________the Unit I pressurizer spray line pipe weld. Section Al1.14 Updated to add fuel tank for portable diesel electric generators credited during fire protection events for _________________supporting safe shutdown. PG&E Letter DCLI15-150 Page 3 of 3 Affected LRA Reason for Change Section Section A1.13 In PG&E Letter DCL-14-1 03, "10 CFR 54.21 (b) Annual Table A4-1, Item 3 Update to the Diablo Canyon Power Plant License Renewal Application (LRA), Amendment 48 and LIRA Appendix E, Applicant's Environmental Report - Operating License Renewal Stage, Amendment 1," dated December 22, 2014, PG&E stated that following the enhancements listed in, Attachment 7C, and with the exceptions listed, the DCPP Fire Water System Program would be consistent with LR-ISG-2012-02, Section C. PG&E has determined that an enhancement not mentioned in DCL-14-103 was necessary for the deluge testing to be consistent with LR-ISG-2012-02, Section C. PG&E is revising the licensing basis to enhance the testing of deluge system nozzles consistent with the 2011 edition of NFPA 25, Section 10.3.4.3.1, as recommended by LR-ISG-2012-02, S Appendix D, Table 4a. Following this enhancement and those listed in DCL-14-1 03, Enclosure 1, Attachment 7C, and with the exceptions listed, PG&Es Fire Water System Program will be consistent with LR-ISG-2012-02, Section C. Table A4-1, Items 34 Updated the status of these items to show them as and 60 completed. Table A4-1, Item 73 of this letter provides DCPP Unit I and 2 responses to the applicable aging management program plant-specific action items, conditions, and limitations identified in the NRC SE, Revision 1, on MRP-227. Table A4-1, item 73 is revised to show this commitment is _________________ complete. PG&E Letter DCL-1 5-1 50 License Renewal Application (LRA) Amendment 51 Affected LRA Sections and Tables, and Figures PG&E Letter DCL-1 5-150 Page 1 of 29 License Renewal Application (LRA) Amendment 51 Affected LRA Sections and Tables, and Figures Table 2.3.3-12 Section 2.4.11 Section 3.1.2.1.2 Section 3.2.2.1.1 Section 3.2.2.1.2 Section 3.2.2.1.3 Table 3.2.2-2 Section 3.3.2.1.3 Section 3.3.2.1.4 Section 3.3.2.1.5 Section 3.3.2.1.8 Section 3.3.2.1.12 Section 3.3.2.1.13 Section 3.3.2.1.14 Section 3.3.2.1.17 Section 3.3.2.1.19 Table 3.3.2-3 Table 3.3.2-8 Table 3.3.2-9 Table 3.3.2-12 Table 3.3.2-14 Section 3.4.2.1.1 Section 3.4.2.1.3 Section 3.4.2.1.4 Section 3.4.2.1.5 Table 3.4.2-1 Section 4.7.5 Section 4.9 Section AI. 13 Section A1.14 Section A3.5.3 Table A4-1, Items 3, 34, 60, and 73 PG&E Letter DCL-1 5-1 50 Page 2 of 29 Table 2.3.3-12 Fire Protection System Section 2.3 SCOPING AND SCREENING RESULTS MECHANICAL SYSTEMS Component Type Intended Function Bellows Pressure Boundary Closure Bolting Pressure Boundary Flow Element Pressure Boundary Flow Indicator Pressure Boundary Hose Station P'ressure..... d.. 'Structural Support Hydrant Pressure Boundary Orifice Pressure Boundary __________________________Throttle Piping Leakage Boundary (spatial) Pressure Boundary __________________________Structural Support Pump Pressure Boundary __________________________Structural Support RCP Oil Collection Reservoir Pressure Boundary Solenoid Valve Pressure Boundary Spray Nozzle Spray Strainer Pressure Boundary Tank Pressure Boundary __________________________Structural Support Test Connection Pressure Boundary Trailer Structural Support Tubing Leakage Boundary (spatial) _________________________Pressure Boundary Valve Leakage Boundary (spatial) Pressure Boundary Vessel Pressure Boundary Section 2.4 PG&E Letter DCL-1 5-150 SCOPING AND SCREENING RESULTS Page 3 of 29 STRUCTURES 2.4.11 Earthwork and Yard Structures Structure Description The earthwork and yard structures include the circulating water conduits, auxiliary saltwater (ASW) vacuum breaker vaults, ASW thrust blocks and anchors, a security-re/a ted enclosure, raw water storage reservoirs 1A and 1 B, east and west breakwaters, and the earth slopes east of the auxiliary building and over the ASW line east of the intake structure. The seismically qualified portions of the circulating water conduits and ASW vacuum breaker vaults are reinforced concrete structures founded on compacted fill. The Design Class I ASW supply piping is supported by reinforced concrete thrust blocks, compacted backfill, and concrete anchors attached to the circulating water conduits. The seismically qualified security-related enclosure is a steel structure and was analyzed to ensure that failure of the enclosure will not impact Design Class I SSCs. The security-related enclosure does not perform any (a)(1) intended functions and does not contain components required by the five License Renewal regulated events (a) (3). The raw water reservoir, located east of the power block, has reinforced concrete-walls. The reservoir is primarily intended to serve as fresh water storage for fire protection and long term cooling. The breakwater structures, which are constructed of precast reinforced concrete blocks and rip-rap, protect the intake structure from tsunami loads. The earth slopes east of auxiliary building and over the ASW line east of the intake structure were analyzed for design basis seismic loads to ensure that such loading will not produce any significant slope failure that can impact Design Class I SSCs. The ASW system buried piping and electrical conduits are protected from tsunami/storm conditions by wave protection measures, which include concrete covers, revetments, roadway slabs, and pavement. Gabion mattresses embedded~within the slopes are covered with grass for additional erosion control. For the purposes of license renewal and aging management, the breakwaters and earth slope protection structures are evaluated as barriers. Structure Intended Functions The earthwork and yard structures, except the security-related enclosure, provide structural support, shelter, and protection for components relied upon to provide the capability to shutdown the reactor and maintain it in a safe shutdown condition. The raw water reservoir provides fresh water storage for long term cooling. The earthwork and yard structures, except the security-related enclosure, also provide structural support, shelter, and protection for nonsafety-related SSCs whose Section 2.4 PG&E Letter DCL-1 5-1 50 SCOPING AND SCREENING RESULTS Page 4 of 29 STRUCTURES failure could prevent performance of a safety-related function. Therefore, the structures are within the scope of license renewal based on the criterion of 10 CFR 54.4(a)(2). The security-related enclosure provides structural support whose failure could prevent performance of a safety-related function. Therefore, the structure is within the scope of license renewal based on the criterion of 10 CFR 54.4(a)(2). The earthwork and yard structures, except the security-related enclosure, provide structural support, shelter, and protection for components required to support fire protection and SBO requirements. Therefore, the earthwork and yard structures are within the scope of license renewal based on the criteria of 10 CFR 54.4(a)(3). Section 3.1 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF REACTOR VESSEL Page 5 of 29 INTERNALS, AND REACTOR COOLANT SYSTEM 3.1.2.1.2 Reactor Coolant System Aging Management Programs The following aging management programs manage the aging effects for the reactor coolant system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) Section 3.2 PG&E Letter DCL-1 5-1 50 AGING MANAGEMENT OF ENGINEERED Page 6 of 29 SAFETY FEATURES 3.2.2.1.1 Safety Injection System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the saltwater and chlorination system component types: Flow-Accelerated Corrosion (B2. 1.6) 3.2.2.1.2 Containment Spray System Materials The materials of construction for the containment spray system component types are: 3.2.2.1.3 Residual Heat Removal System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the saltwater and chlorination system component types: Flow-Accelerated Corrosion (B2.1.6) PG&E Letter DCL-15-150 Page 7 of 29 Section 3.2 AGING MANAGEMENT OF ENGINEERED SAFETY FEATURES Table 3. 2.2-2 Engineered Safety Features - Summary of Aging M anagement Evaluation - Containment Spra y System Component Intended Material Environment Aging Effect Aging Management NUREG-Table I Item Notes Type Function Requiring Program 1801 Vol. Management 2 Item VaIve R-B G*,n"f-A (!nt) No None V--F--4 3*~4-5 A Valve R-B Gop-lo PatIdo Ai Non Ne* V-F-A Section 3.3 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 8 of 29 3.3.2.1.3 Saltwater and Chlorination System Materials The materials of construction for the saltwater and chlorination system component types are: Ductile Iron Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the saltwater and chlorination system component types: Flow-Accelerated Corrosion (B2. 1.6) Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.3.2.1.4 Component Cooling Water System Aging Management Programs The following aging management programs manage the aging effects for the component cooling water system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2.1. 42) 3.3.2.1.5 Makeup Water System Aging Management Programs The following aging management programs manage the aging effects for the makeup water system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (82.1.42) Section 3.3 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 9 of 29 3.3.2.1.8 Chemical and Volume Control System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the chemical and volume control system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1. 42) 3.3.2.1.12 Fire Protection System Aging Management Programs The following aging management programs manage the aging effects for the fire protection system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.3.2.1.13 Diesel Generator Fuel Oil System Aging Management Programs The following aging management programs manage the aging effects for the diesel generator fuel oil system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components. Heat Exchangers, and Tanks (B2. 1.42) 3.3.2.1.14 Diesel Generator System Materials The materials of construction for the diesel generator system component types are: Copper Alloy (> 15 percent Zinc) Section 3.3 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 10 of 29 3.3.2.1.17 Liquid Radwaste System Aging Management Programs The following aging management programs manage the aging effects for the liquid radwaste system component types: S Internal Coatings/Linings for in-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.3.2.1.19 Oily Water and Turbine Sump System Aging Management Programs The following aging management programs manage the aging effects for the oily water and turbine sump system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 142) PG&E Letter DCL-1 5-1 50 Page 11 oft29 SECTION 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3. 3.2-3 Auxiliary Systems - Summary of Aging Management Evaluation - Saltwater and Chlorination System Component Intended Material Environment Aging Effect Aging Management NUREG-Table 1 Item Notes Type Function Requiring Program 1801 Vol. _______Management 2 Item Separator LBS Nickel Alloys Plant Indoor Air None None VII.J-14 3.3.1.94 A ~(Ext) Separator LBS Nickel Alloys Raw Water (lnt) Loss of material Open-Cycle Cooling VII. C1-13 3.3.1.78 A Water System (B2. 1.9) Valve LBS Ductile Iron Plant Indoor Air Loss of material External Surfaces VIII-8 3.3.1.58 B (Ext) Monitoring Program (B2. 1.20) Valve LBS Ductile Iron Raw Water (Int) Loss of material Open-Cycle Cooling VII. C1-19 3.3.1.76 A Water System (B2. 1.9) Valve LBS Ductile Iron Raw Water (Int) Wall thinning due Flow-Accelerated None None H, 5 ____to erosion Corrosion (B2. 1.6)_ PG&E Letter DCL-15-150 Page 12 of 29 SECTION 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS T~hlk R.* 2-R A~iviliart, 5Zv~efmjc -- 5Zlimmnrv nf Atninn AAn~ninmimnt ~I::inhIi~tin -- flht.mir'l ~nnr I l/AIm* Crinfrrl.VIQfctim Component Intended Material Environment Aging Effect Aging Management Program NUREG-Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item Piping FB, LBS, Carbon Plant Indoor Air Loss of material External Surfaces Monitoring VII.I-8 3.3.1.58 8, 12 PB, SIA Steel Ex)Program (B2.1.20) Piping FB, LBS, Carbon Plant Indoor Air Loss of material Inspection of Internal Surfaces V.A-19 3.2.1.32 B3, 12 SIA Steel (Int) in Miscellaneous Piping and Ducting Components (B2.1.22) Notes for Table 3.3.2-8: Plant Specific Notes: 12 The guard pipe enclosing the hydrogen piping is credited in the Fire Hazards Analysis and is thus in the scope of License Renewal per 10 CFR 54. 4(a)(3), The guard pipe is not a fire barrier since it does not prevent the spread of fire, as defined in Regulatory Guide 1.120. The most-appropriate and most-conservative license renewal intended function of "fire barrier" was chosen to support its inclusion as being in-scope because it is relied upon in the Fire Hazards Analysis. Table 3.3.2-9 Auxiliary Systems - Summary of Aging Management Evaluation - Miscellaneous HVAC Systems Component Intended Material Environment Aging Effect Aging Management NUREG-Table I Item Notes Type Function Requiring Program 1801 Vol. Management 2 Item Valve SIA, SS Copper Alloy Plant Indoor Air None None VIII.I-2 3.4.1.41 C ________Ext) Valve SIA, SS Copper Alloy Ventilation Loss of material Inspection of Internal VII.G-9 3.3.1.28 iE Atmosphere (Int) Surfaces in Miscellaneous Piping and Ducting _________Components_(B2.1.22) PG&E Letter DCL-15-150 Page 13 of 29 SECTION 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-12 Auxiliary Systems - Summary of Aging Management Evaluation - Fire Protection System Component Intended Material Environment Aging Effect Aging Management Program NUREG-Table I Notes Type Function Requiring 1801 Vol. Item Management 2 Item Hose Station P-BSS Carbon Steel Atmosphere/ Loss of material External Surfaces Monitoring VII.I-9 3.3.1.58 B, 16 Weather (Ext) Program (B2. 1.20) Hose Station P-BSS Carbon Steel Plant Indoor Air Loss of material External Surfaces Monitoring VII.I-8 3.3.1.58 B, 16 (Ext) Program (B2.1.20) Hose-Station P-B Gabn-te R ae-{* Loss ofmateria! Fire Water Sys*tem, (B2.1-1-3 V-G). 34-448S B Hose-Station P-B Cabn-te Ra-aer-Tt ruroenl Fire Water System (B2.!.13-) None None Gorrosio

OiiePB, TH Copper Alloy Atmosphere!

Loss of material Selective Leaching of None None G Oiie(> 15% Zinc) Weather (Ext Materials (B2. 1.17)

OiiePB, TH Copper Alloy Plant Indoor Air None None VIII.I-2 3.4.1.41 A

Oriice(> 15% Zinc) Ext

OiiePB, TH Copper Alloy Plant Indoor Air Loss of material Fire Water System (B2. 1.13)

VlI.G-9 3.3.1.28 E, 8 Oriice(> 15% Zinc) (Int Piping PB Copper Alloy Atmosphere! Loss of material Selective Leaching of None None G "> 15% Zinc) Weather (Ext Materials (B2. 1.17) Piping PB Copper Alloy Plant Indoor Air None None VIII.I-.2 3.4.1.41 A '> 15% Zinc) ESxt) PB Copper Alloy Plant Indoor Air Loss of material Fire Water System (B2. 1.13) VII.G-9 3.3.1.28 E, 8 Piping (> 15% Zinc) (Int) Spray Nozzle SP Copper Alloy Plant Indoor Air Loss of Fire Water System VII. G-3.3.1.28Nort E, 8G (_______> 15% Zinc) (Int) materialNepne (B2. 1. 13)Nene 9Nene-e SECTION 3.3 PG&E Letter DCL-15-150 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 14 of 29 Notes for Table 3.3.2-12: Plant Specific Notes: 8 The Fire Water System program (B2.1 13) is used to monitor copper alloy piping, piping components and piping elements exposed to condensation (internal) for loss of material in the fire protection system. Reference LR-ISG-2012-02, Appendix C, Line VII.G.A-1 43, aind PG&E Letter DCL-14-1 03, Enclosure 1, Attachment 7C, and DCL-15-150, Enclosure 2.

16.

This line item only represents the hose reel and not the associated isolation valve, piping, or fittings. PG&E Letter DCL-15-150 Page 15 of 29 SECTION 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-14 Auxiliary Systems - Summary of Aging Management Evaluation - Diesel Generator System Component Intended Material Environment Aging Effect Aging Management NUREG-Table I Item Notes Type Function Requiring Program 1801 Vol. ______Management 2 Item_____ Valve PB Aluminum Plant Indoor Air Loss of material Inspection of Internal VII.F2-12 3.3. 1.27 E (Int) Surfaces in Miscellaneous Piping and Ducting Components (B2. 1.22) Valve PB Aluminum Plant Indoor Air None None VII. J-1 3.3.1.95 A (Ext) Valve PB Copper Alloy Lubricating Oil Loss of material Lubricating Oil Analysis VII. H2-10 3.3.1.26 B (> 15% Zinc) (Int) (B2. 1.23) and One-Time Inspection (B2. 1.16) Valve PB Copper Alloy Plant Indoor Air None None V.F-3 3.2.1.53 A ______(> 15% Zinc) (Ext)_______ Section 3.4 PG&E Letter DCL-1 5-150 AGING MANAGEMENT OF STEAM AND Page 16 of 29 POWER CONVERSION SYSTEM 3.4.2.1.1 Turbine Steam Supply System Aging Management Programs The following aging management programs manage the aging effects for the turbine steam supply system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.4.2.1.3 Feedwater System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion 3.4.2.1.4 Condensate System Aging Management Programs The following aging management programs manage the aging effects for the condensate system component types: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (B2. 1.42) 3.4.2.1.5 Auxiliary Feedwater System Aging Effects Requiring Management The following saltwater and chlorination system aging effects require management: Wall thinning due to erosion Aging Management Programs The following aging management programs manage the aging effects for the saltwater and chlorination system component types: S Flow-Accelerated Corrosion (B2. 1.6) PG&E Letter DCL-1 5-1 50 Page 17 of 29 SECTION 3.4 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Table 3.4.2-1 Steam and Power Conversion System - Summar'y of Aging Management Evaluation - Turbine Steam Sunoltv System Component Intended Material Environment Aging Effect Aging Management NUREG-Table I Item Notes Type Function Requiring Program 1801 Vol. Management 2 Item Expansion LBS Nickel Alloys Plant Indoor Air None None VIII.I-9 3.4.1.41 A Joint (Ext) Expansion LBS Nickel Alloys Steam (Int) Loss of material Water Chemistry VIII. B1-1 3.4.1.37 E, 4 Joint (B2. 1.2) and One-Time ____ Inspection (B2. 1.16) Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 18 of 29 4.7.5 Inservice Flaw Growth Analyses that Demonstrate Structural Stability for 40 Years Summary Description The 1SI procedure states that a fracture mechanics analysis, in accordance with ASME Section XI Code, Subsection IWB-3600, must be completed if a flaw acceptance criterion is not met as outlined in the corresponding test procedure. These analyses depend on a specified number of operating years, and thus may be TLAAs for DCPP. Analysis Unit 2 RHR Piping Weld RB-119.11 During a routine inservice inspection prior to DCPP Unit 2 Refueling Outage 13 (2R1 3) in 2006, a circumferential flaw was identified in Weld RB-I119-11 of the residual heat removal (RHR) system. The observed flaw did not meet the Section Xl acceptance standards of Table IWB-3514-2. Consequently, the indication was evaluated per the guidelines of Section Xl, IWB-3640. A conservative fatigue crack growth evaluation was then performed to determine the adequacy of the piping system for continued operation. The evaluation was submitted to the NRC for review, as required by the Code, in PG&E Letter DCL-06-069. The service life for Weld RB-i119-11 is based on operating for 40 years from the date the flaw was identified, i.e. until 2046, during which the flaw would experience 500 startup-shutdown cycles. Thus, the evaluation encompassed a 60-year plant life and the analysis will be valid beyond the 2045 end date of the period of extended operation for Unit 2. The cycle assumptions used in the analysis are conservative compared to the DCPP original design cycles described in Section 4.3.1.1. The DCPP licensing basis assumes 250 heatups and 250 cooldowns for a 50 year plant life. Since the analysis indicates that the allowable flaw depth will not be reached for the next 40 years of plant operation beginning in October 2006, the flaw evaluation of RHR Weld RB-i119-11 will remain valid for the period of extended operation in accordance with 10 CFR 54.211l(1)(i). Unit 2 Auxiliary Feedwater Piping Line 567 During Unit 2 Refueling Outage 8 (2R8), while performing a non-routine surface examination prior to maintenance, DCPP identified a flaw indication in the auxiliary feedwater pump recirculation header Line 567, that exceeds Section Xl, Table IWB-3410-1 criteria. The flaw has been accepted by analysis by meeting the Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 19 of 29 allowable size criteria of IWB-3620 and IWB-3610 and was submitted to the NRC in PG&E Letter DCL-99-1 36. The numbers of thermal and seismic cycles used in the analysis are consistent with or more conservative than the DCPP 50-year design basis described in FSAR Table 5.2-4. The assumed transients are consistent with or bounded by the 50 year licensing basis. The number of transients will be monitored by the enhanced Fatigue Management Program. The enhanced Fatigue Management Program provides assurance that the fatigue crack growth analysis will be managed for the period of extended operation in accordance with 10 CFR 54.211(1 )(iii). Unit I RHR Piping Weld WIC-95 During Unit 1 Refueling Outage 9 (1R9), while performing an inservice inspection, DCPP identified a weld flaw indication located in an ASME Class 2 portion of the RHR injection Line 985 to hot legs 1 and 2 at weld WLC-95. The indication exceeded the Section Xl, Table IWC-3410-1 criteria. The flaw has been accepted by analysis in accordance with IWB-3410 and was submitted to the NRC in PG&E Letter DCL-97-086. The number of seismic cycles used in the analysis is consistent with the DCPP 50-year design basis described in FSAR Table 5.2-4. There have been no occurrences of a DE, DDE, or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the Unit 1 RHR Weld WIC-95 fatigue crack growth evaluation for the 50-year design basis number of DE, DDE, and Hosgri events is sufficient to the end of the period of extended operation. Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CFR 54.211l(1)(i). Units I and 2 Safety Injection Pumps Vent and Drain Socket Welds In 2014, PG&E requested NRC approval of lnservice Inspection Request for Alternative REP-SI for DCPP, Units 1 and 2. To support this request, a stress and fracture mechanics evaluation was performed to determine the adequacy of the socket welds associated with ASME Class 2 Safety Injection (SI) Pumps 1-1, 1-2, and 2-1 vent and drain connections. The evaluation was submitted to the NRC for review, as part of the relief request, in PG&E Letter DCL-14-060, dated July 21, 2014. Postulated flaws were evaluated using a fracture mechanics approach analogous to the methods of ASME Code Section XI. This relief request was approved for the remaining life of the subject SI Pumps, including the duration of the current operating licenses plus a license renewal period of 20 years [Reference 42]. Since the evaluation is based on the 60-year operating period, the TLAA covers the period of extended operation and is dispositioned under 10 CFR 54.2 11(1)(i). Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 20 of 29 For the postulated crack analysis, 7, 000 thermal transient cycles (pump starts), 400 DE cycles (20 events with 20 cycles per event), and 20 Hosgri earthquake cycles (1 event with 20 cycles) were assumed. Using a conservative projection of 1,400 SI Pump start cycles for a 60 year plant life, the 7, 000 thermal transient cycles assumed in the postulated crack analysis during 60 years of operation is conservative. The number of seismic cycles used in the analysis is consistent with the DCPP 50-year design basis described in FSAR Table 5.2-4. There have been no occurrences of a DE or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the postulated crack analysis for the 50-year design basis number of DE and Hosgri events is sufficient to the end of the period of extended operation. The analysis is valid for the period of extended operation in accordance with 10 CFR 54.211(1)(i). Unit 2 Pressurizer Safety and Spray Nozzle Welds As stated in LRA Section 4. 7.2, during Unit 2 Refueling Outage 14 (2R 14, Spring 2008), Alloy 690 structural weld overlays were completed on Alloy 82/182 welds attaching the surge, spray, and relief valve nozzles to the safe ends, and the safe ends to the connecting piping. During the seventeenth Unit 2 Refueling Outage (2R1 7), while performing inservice inspections, DCPP identified weld flaw indications located at Unit 2structural weld overlays for the pressurizer safety nozzles Aand B, and pressurizer spray nozzle. Conservative fatigue crack growth evaluations were then performed to determine the adequacy of the piping system for continued operation. The evaluations were submitted to the NRC for review, as part of a relief request, in PG&E Letter DCL 028, dated April 7, 2014. This relief request was approved for the service life of the structural weld overlays [Reference 43]. The service life for the pressurizer safety and spray nozzle structural weld overlays is based on operating for 38 years from the date the structural weld overlays were completed, i e. until 2046. Thus, the evaluation encompassed a 60-year plant life and the analysis will be valid beyond the 2045 end date of the period of extended operation for Unit 2. The cycle assumptions used in the analyses are consistent with those transients used in the pressurizer structural weld overlay (LRA Section 4.7.2 and PG&E Letter DCL-10-120). Per LRA Table A 4-1, Commitment 38, the plant transient cycles related to the structural weld overlay fatigue crack growth analyses are included in the existing plant transient monitoring program. Since the analyses indicate that the allowable flaw depth will not be reached for the remaining plant life, the flaw evaluations of the pressurizer safety and spray nozzle structural weld overlays will remain valid for the period of extended operation in accordance with 10 CFR 54.211(1)(i). Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 21 of 29 Unit I Pressurizer Spray Line Pipe Weld WIB-378 During Unit 1 Refueling Outage 19 (1R 19), while performing an inservice inspection, DCPP identified a weld flaw indication located in an ASME Code Class 1 pressurizer spray line pipe weld WlB-3 78. The indication exceeded the Section Xl, Table IWB-35 14-2 criteria. The flaw has been accepted by analysis in accordance with iWB-3600 and was submitted to the NRC in PG&E Letter DCL-15-131, dated November 3, 2015. A fatigue crack growth evaluation was performed to determine the adequacy of the weld for continued operation. The indication was a planar flaw oriented circumferentially and was assumed to be ID connected for conservatism. The service life for Weld WIB-378 is based on operating through the period of extended operation. Thus, the evaluation encompassed a 60-year plant life and the analysis will remain valid for the Unit 1 period of extended operation. The number of transient cycles used in the analysis is consistent with or more conservative than the DCPP 50-year design basis described in FSAR Table 5.2-4. Because the evaluation indicates that the allowable flaw depth will not be reached for the remaining plant life and the assumed transients are consistent with or more conservative than the DCPP 50-year design basis described in FSAR Table 5. 2-4, the flaw evaluation of the pressurizer spray line pipe weld WlB-378 will remain valid for the period of extended operation in accordance with 10 CFFR 54.2 11(1)(i). Disposition: Validation, 10 CFR 54.21 1(1 )(i); and Aging Management, 10 CFR 54.211l(1)(iii) Validation - Flaw Evaluation of Unit 2 RHR Piping Weld RB-119-II The result indicates that the allowable flaw depth will not be reached for the next 40 years of plant operation beginning in October 2006. Therefore, the flaw evaluation of RHR Weld RB-I119-11 will remain valid for the period of extended operation in accordance with 10 CFR 54.21 I(1)(i). Validation - Flaw Evaluation of Unit I RHR Weld WIC-95 There have been no occurrences of a DE, DDE, or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the Unit 1 RHR Weld WIC-95 fatigue crack growth evaluation for the 50-year design basis number of DE, DDE, and Hosgri events is sufficient to the end of the period of extended operation. Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CER 54.211(1)(i). Section 4 PG&E Letter DCL-1 5-1 50 TIME-LIMITED AGING ANALYSES Page 22 of 29 Validation - Units 1 and 2 Safety Injection Pumps Vent and Drain Socket Welds Using a conservative projection of 1,400 SI Pump start cycles for a 60 year plant life, the 7, 000 thermal transient cycles assumed in the postulated crack analysis during 60 years of operation is conservative. There have been no occurrences of a DE or Hosgri seismic event at DCPP during the first 20 plus years of operation. Therefore, the seismic cycles in the postulated crack analysis for the 50-year design basis number of DE and Hosgri events is sufficient to the end of the period of extended operation. The analysis is valid for the period of extended operation in accordance with 10 CFR~ 54.211(1)(i). Validation - Flaw Evaluation of Unit 2 Pressurizer Safety and Spray Nozzle Welds The results indicate that the allowable flaw depth will not be reached for the remaining plant life. Therefore, the flaw evaluation of Unit 2 Pressurizer Safety and Spray Nozzle Welds will remain valid for the period of extended operation in accordance with 10 CFR 54.2 11(1)(i). Validation - Flaw Evaluation of Unit I Pressurizer Spray Line Pipe Weld WIB-378 The results indicate that the allowable flaw depth will not be reached for the remaining plant life. Therefore, the flaw evaluation of Unit 1 Pressurizer Spray Line Pipe Weld W/B-378 will remain valid for the period of extended operation in accordance with 10 CFR 54.211(1)(i). Aging Management - Unit 2 Auxiliary Feedwater Piping Line 567 The Metal Fatigue of the Reactor Coolant Pressure Boundary program (B3. 1) monitors fatigue design transients including the transients assumed in the fatigue crack growth analyses for the Unit 2 auxiliary feedwater piping Line 567. The program provides assurance that the fatigue crack growth analysis will be managed for the period of extended operation in accordance with 10 CFR 54.211(1 )(iii). Section 4 PG&E Letter DCL-15-150 TIME-LIMITED AGING ANALYSES Page 23 of 29

4.9 REFERENCES

42.

US NRC Letter. From Michael T. Markley, Chief, Plant Licensing Branch IV-l, Division of Operating Reactor Licensing, Off~ice of Nuclear Reactor Regulation; to Mr. Edward D. Halpin, Senior Vice President and Chief Nuclear Officer, DCPP. "Diablo Canyon Power Plant, Unit Nos. 1 and 2 - Request for Alternative REP-SI, Revision 2, Proposed Alternative to Requirements for Repair/Replacement Activities for Certain Safety Injection Pump Welded Attachments (TAC Nos. MF4476 and MF44 77)." 15 July 2015. (ML15187A035)

43.

US NRC Letter. From Michael T. Markley, Chief, Plant Licensing Branch IV-1, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation; to Mr. Edward D. Halpin, Senior Vice President and Chief Nuclear Officer, DCPP. "Diablo Canyon Power Plant, Unit No. 2 - Inservice Inspection Program Relief Request SWOL-REP-1 U2 for Approval of an Alternative to the ASME Code, Section XI, for Preemptive Full Structural Weld Overlays (TAC No. MF389 1)." 14 October 2014. (ML14255A232) Appendix A PG&E Letter DCL-1 5-150 Final Safety Analysis Report Supplement Page 24 of 29 AI. 13 FIRE WATER SYSTEM The Fire Water System program manages loss of material due to corrosion, including MIC, fouling, flow blockage because of fouling, and loss of integrity for water-based fire protection systems and internal coatings/linings for the fire water storage tank within the scope of license renewal. Internal and external inspections and tests of fire protection equipment are performed consistent, with exceptions identified in PG&E Letters DCL-14-1 03, Enclosure 1, Attachment 7C, and DCL-15-121 with NFPA-25 (2011 edition). Testing or replacement of sprinklers that have been in place for 50 years is performed in accordance with NFPA-25 (2011 edition). Portions of the deluge systems that are normally dry but periodically subjected to flow and cannot be drained or allow water to collect will undergo augmented testing beyond that in NFPA-25 consisting of volumetric wall thickness examinations. The fire water system is managed by performing routine preventive maintenance, inspections and testing; operator rounds, performance monitoring, and reliance on the corrective action program; and system improvements to address aging and obsolescence issues. The fire water system is normally maintained at required operating pressure and is monitored such that loss of system pressure is immediately detected and corrective actions are initiated. The Fire Water System program will conduct a flow test with air, water, or other medium through each open spray nozzle to verify that deluge systems nozzles are unobstructed. Water flow tests will verify that the deluge system provide full coverage of the equipment it protects. Visual inspections will be performed on firewater piping. Non-intrusive follow-up volumetric examinations will be performed if internal visual inspections detect surface irregularities to determine if wall thickness is within acceptable limits. Visual inspections will evaluate for the presence of sufficient foreign material to obstruct fire water pipe or sprinklers. Inspections of the firewater tank will be performed to detect loss of material. As discussed in PG&E Letter DCL-1 5-027, Enclosure 1, in response to LR-ISG-2013-01, the program consists of periodic visual inspections of the internal liner of the fire water storage tank exposed to raw water where loss of lining integrity could impact the components' and downstream components' current licensing basis intended function(s). For coated surfaces determined to not meet the acceptance criteria, physical testing is performed where physically possible (i.e., sufficient room to conduct testing) in conjunction with repair, replacement, or removal of the lining. The training and qualification of individuals involved in coating inspections are conducted in accordance with ASTM International Standards endorsed in RG 1.54 including guidance from the NRC associated with a particular standard. The Fire Water program implements the recommendations in LR-ISG-2012-02, as discussed in PG&E Letters DCL-14-1 03, Enclosure 1, Attachments 7C,-arn4 DCL-15-121, and DCL-15-150, the recommendations in LR-ISG-2013-01, as discussed in PG&E Letter DCL-1 5-027 Appendix A PG&E Letter DCL-1 5-150 Final Safety Analysis Report Supplement Page 25 of 29 A1.14 FUEL OIL CHEMISTRY The Fuel Oil Chemistry program manages loss of material on the internal surface of components in the emergency diesel fuel oil storage and transfer system, portable diesel electric generator fuel oil tanks, portable diesel driven fire pump fuel oil tanks, and portable caddy fuel oil tanks. The program includes (a) surveillance and monitoring procedures for maintaining fuel oil quality by controlling contaminants in accordance with applicable ASTM Standards, (b) periodic draining of water from fuel oil tanks, (c) visual inspection of internal surfaces during periodic draining and cleaning, (d) one-time ultrasonic wall thickness measurements of accessible portions of fuel oil tank bottoms, (e) sampling and analysis of new fuel oil before it is introduced into the fuel oil tanks, and (f) supplemental one-time inspections of a representative sample of components in systems that contain fuel oil by the One-Time Inspection program (A1.16). Appendix A PG&E Letter DCL-1 5-150 Final Safety Analysis Report Supplement Page 26 of 29 A3.5.3 Inservice Flaw Growth Analyses that Demonstrate Structural Stability for 40 Years The ISI procedure states that a fracture mechanics analysis, in accordance with ASME Code, Section Xl, Subsection IWB-3600, must be completed if flaw acceptance criterion is not met as outlined in the corresponding test procedure. These analyses depend on a specified number of operating years, and thus may be TLAAs. Unit 2 RHR Piping Weld RB-119-11 In 2006, a circumferential flaw was identified in DCPP Unit 2 Weld RB-i119-11 of the residual heat removal (RHR) system. The observed flaw did not meet the Section XI acceptance standards of Table IWB-3514-2. Consequently, the indication was evaluated per the guidelines of Section Xl, IWB-3640. A conservative fatigue crack growth evaluation was performed to determine the adequacy of continued operation of the piping system. The analysis is based on operating for 40 years from the date the flaw was identified and will be valid beyond the end of the period of extended operation for Unit 2 in accordance with 10 CFR 54.21 1(1 )(i). Unit 2 Auxiliary Feedwater Piping Line 567 DCPP identified a flaw indication in the Unit 2 auxiliary feedwater pumps recirculation header Line 567, that exceeds Section Xl, Table IWB-3410-1 criteria. The flaw has been accepted by analysis by meeting the allowable size criteria of IWB-3620 and IWB-3610. The Metal Fatigue of Reactor Coolant Pressure Boundary program described in Section A2.1I monitors fatigue design transients including the transients assumed in the fatigue crack growth analyses and therefore will be managed for the period of extended operation in accordance with 10 CFR 54.211l(1)(iii). Unit I RHR Piping Weld WIC-95 DCPP identified a weld flaw indication located in an ASME Class 2 portion of the Unit 1 residual heat removal injection Line 985 to hot legs 1 and 2 at weld WIC-95. The indication exceeded the Section Xl, Table IWC-341 0-1 criteria. The flaw has been accepted by analysis in accordance with IWB-3410. The number of seismic cycles assumed in the analysis is sufficient for the period of extended operation. Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CFR 54.211(1)(i). Appendix A PG&E Letter DCL-1 5-150 Final Safety Analysis Report Supplement Page 27 of 29 Units 1 and 2 Safety Injection Pumps Vent and Drain Socket Welds In support of a relief request, a stress and fracture mechanics evaluation was performed to determine the adequacy of socket welds associated with ASME Class 2 Safety Injection (SI) Pumps 1-1, 1-2, and 2-1 vent and drain connections. Postulated flaws were evaluated using a fracture mechanics approach analogous to the methods of ASME Code Section Xl. The number of cycles used in the analysis is sufficient for the period of extended operation. Using a conservative projection of 1,400 SI Pump start cycles for a 60 year plant life, the 7, 000 thermal transient cycles assumed in the postulated crack analysis during 60 years of operation is conservative. Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CFR 54.2 11(1)(i). Unit 2 Pressurizer Safety and Spray Nozzle Welds In 2013, laminar flaws were identified in DCPP Unit 2 structural weld overlays for pressurizer safety nozzles A and B, and pressurizer spray nozzle. Conservative fatigue crack growth evaluations were performed to determine the adequacy of continued operation of the piping system. The analyses are based on operating for 38 years from the date the structural weld overlays were completed and will be valid beyond the end of the period of extended operation for Unit 2 in accordance with 10 CFR 54.2 11(1)(i). Unit 1 Pressurizer Spray Line Pipe Weld WIB-378 In 2015, a circumferential flaw was identified in DCPP Unit 1 Pressurizer Spray Line Pipe Weld WIB-3 78. The observed flaw did not meet the Section Xl acceptance standards of Table IWB-3514-2. Consequently, the indication was evaluated per the guidelines of Section XI, IWB-3600. A conservative fatigue crack growth evaluation was performed to determine the adequacy of continued operation of the wveld. The analysis is based on operating through the period of extended operation and will remain valid Unit 1 in accordance with 10 CFR 54.211(1)(i). PG&E Letter DCL-15-150 Page 28 of 29 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Table A4-1I License Renewal Commitments LRA Implementation Item # Commitment Section Schedule 3 Enhance the Fire Water System program: B2.1.13 Program is (a) Sprinkler heads in service for 50 years will be replaced or representative samples from implemented 5 one or more sample areas will be tested consistent with NFPA 25, Inspection, Testing years before the and Maintenance of Water-Based Fire Protection Systems, 2011 Edition guidance. period of extended Test procedures will be repeated at 10-year intervals during the period of extended operation. operation, for sprinkler heads that were not replaced prior to being in service for 50 Inspections of years, to ensure that signs of degradation, such as corrosion, are detected prior to the wetted normally loss of intended function, and dry piping (b) To perform non-intrusive follow-up volumetric examinations if internal visual segments that inspections detect surface irregularities to determine if wall thickness is within cannot be drained acceptable limits. Visual inspections will evaluate for the presence of sufficient foreign or that allow water material to obstruct fire water pipe or sprinklers to collect begin 5 (c) To be in conformance with LR-ISG-2012-02, Section C as discussed in PG&E Letter years before the DCL-14-1 03, Enclosure 1, Attachment 7C. period of extended (d) To be in conformance with LR-ISG-2013-01 as discussed in PG&E Letter DCL-15-027, operation. Internal. linings inspections (e) Test deluge system nozzles in accordance with the 2011 Edition of NFPA 25, begin no later than Section 10.3.4.3. 1. the last refueling outage before the period of extended operation. The program's remaining inspections begin during the period of extended __________ operation 34 The DCPP work control procedure will be revised to include evaluation of reinforced concrete B1.232 Complete. PG&E exposed during excavations. Letter DCL-15-150 PG&E Letter DCL-1 5-1 50 Page 29 of 29 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Table A4-1 License Renewal Commitments LRA Implementation Item # Commitment Section Schedule 60 PG&E will enhance provisions in the HVAC ducting from the 480V switchgear room that allow water to drain from the exhaust ducting so water cannot enter the 480V switchgear room. Complete. PG&E Letter DCL 150..PfotG4 The NRC SE for MRP-227 contains eight action items for applicants/licensees to consider. Cmlt.P& Responses to the applicable aging management program plant-specific action items, Comp.41letter PG&E5-50 73 conditions, and limitations identified in the NRC SE, Revision 1, on MRP-227 will be submitted D214 ecemerDC2015-10 to the NRC by December 2015. Reference DCL-14-1 03, Enclosure 1, Attachment 4. PG&E Letter DCL-1 5-1 50 WCAP-1 7462-NP, Revision 1 Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon Power Plant Unit I PG&E Letter DCL-1 5-1 50 Page 1 of 5 MRP-227-A Applicability Guideline for Diablo Canyon Power Plant Westinghouse Pressurized Water Reactor Design

Background

The Nuclear Regulatory Commission (NRC) staff has determined that additional information, as discussed in References 1 and 2, should be provided by licensees to verify the applicability of MRP-227-A (Reference 6). The two specific generic issues that need to be addressed are summarized as follows:

1. Do the reactor internals have any non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater, and if so, do the affected components have operating stresses greater than 30 ksi?
2. Does the plant have atypical fuel design or fuel management that could render the assumptions of MRP-227-A, regarding core loading/core design, non-representative?

PG&E Response to Question I Diablo Canyon Power Plant (DCPP) Units 1 and 2 reactor internals components have been evaluated according to industry guideline MRP 2013-025 (Reference 3), as well as to the MRP-1 91 (Reference 4) industry generic component listings and screening criteria (including consideration of cold work as defined in MRP-175 (Reference 5), noting the requirements of Section 3.2.3). In addition to consideration of the material fabrication, forming, and finishing process, a general screening definition of "severe cold work" [a resulting reduction in wall thickness (material stock thickness) of 20 percent] was applied as an evaluation limit. The evaluation included a review of all plant modifications affecting reactor internals and the plant operating history. The components were procured according to American Society for Testing and Materials International of American Society of Mechanical Engineers material specifications that were callouts in the original plant construction drawings. Thus, material identification based on the material callouts and notes in the component drawings was an efficient and reasonable approach to identify the material of construction of components for DCPP Units 1 and 2. Based on the specifications used in the DCPP Units land 2 plant component drawings, it was possible to bin the reactor internals components into five material categories identified in MRP 2013-025. DCPP Units I and 2 components were binned according to the following categories for the materials used in the component fabrication. PG&E Letter DCL-1 5-1 50 Page 2 of 5 Categories based on MRP 2013-025 include:

  • Cast austenitic stainless steel (CASS) (Category 1)
  • Hot-formed austenitic stainless steel (Category 2)
  • Annealed austenitic stainless steel (Category 3)
  • Fasteners austenitic stainless steel (Category 4)
  • Cold-formed austenitic stainless steel without subsequent solution annealing (Category 5)

The potential for cold work is directly controlled by the materials specifications. Essentially, all of the components that are binned (based on their specified materials) as Categories 1, 2, and 3 are non-cold worked; therefore, they have less than 20 percent cold work according to NRC criterion. Similarly, any component binned under Category 5 has the potential to contain greater than 20 percent cold work. Category 4 materials are fasteners that may have been intentionally strain-hardened. The strain hardening according to guidelines should have been intentionally restricted to less than 20 percent. Material definitions in drawings identify maximum yield stress restrictions on these materials, which allows for the identification of the cold work level. In some cases, however, these restrictions are not present on drawings. Restrictions or limitations on the material yield stress (e.g., a maximum of 90 ksi) would indicate that the material cold work would be limited to less than 20 percent. In the absence of a maximum restriction yield stress of strain-hardened material, a conservative approach was taken to indicate the potential for greater than 20 percent cold work. Where multiple options existed for a component or assembly, the bounding condition was taken as the option that had the greater potential to include greater than 20 percent cold work. This. option was then employed in the assessment of the component and was selected for the purposes of the Westinghouse evaluation. In some instances, sequential fabrication would appear to mitigate any potential for cold work; however, since the historical record was not detailed, the potential is noted, but a conservative approach was selected for the Westinghouse evaluation. The evaluation, performed consistently with MRP 2013-025, concluded that the reactor internals Categories 1, 2, and 3 (non-bolting) components at DCPP Units I and 2 contain no cold work greater than 20 percent as a result of material specification and controlled fabrication construction. Category 4 components were already assumed to have the potential for coldwork in the MRP-191 generic assessments. No Category 5 components with severe cold work were identified for DCPP Units I and 2. The detailed evaluation for~the DCPP Units 1 and 2 cold work assessments concluded that the plant-specific fabrication and design was consistent with the PG&E Letter DCL-15-150 Page 3 of 5 MRP-1 91 basis, and that the MRP-227-A (Reference 6) sampling inspection aging management requirements, as related to cold work, are directly applicable to DCPP Units I and 2. The inspection sampling requirements for aging management outlined in MRP-232 (Reference 7) are based on the assumptions of MRP-191 and MRP-227-A. Therefore, MRP-232 calls for the demonstration that the plant specific materials, fabrication, and design meet the assumptions inherent in MRP-191 and MRP-227-A. The detailed Westinghouse evaluation of DCPP Units I and 2 material fabrication and design has concluded that no non-fastener materials of greater than 20 percent cold work were used in construction, and that the inspection sampling approach of MRP-232 is applicable to DCPP Units I and 2. PG&E Response to Question 2 As stated in MRP 2013-025, to demonstrate plant-specific applicability of the MRP-227-A sampling inspection strategy for managing aging in reactor internals, licen~sees must demonstrate that the criteria of MRP-227-A, Section 2.4 are met, and that the neutron fluence and heat generation rates are within the range of the following variables summarized. As detailed in WCAP-1 7462-NP, Revision 1, "Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon Power Plant Unit 1," and WCAP-17463-NP, Revision 1, "Program Plan for Aging Management of Reactor Vessel Internals at Diablo Canyon Power Plant Unit 2," for DCPP Units I and 2, respectively, the criteria specified in MRP-227-A, Section 2.4 has been demonstrated as follows. The MRP-227-A, Section 2.4 assumptions are stated first, followed by' a description of how the assumptions are addressed at DCPP Units I and 2. The assumptions from MRP-227-A, Section 2.4 are as follows:

  • 30 years of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low leakage fuel management strategy for the remaining 30 years of operation; DCPP Units 1 and 2 fuel management programs changed from a high-to
  • low-leakage core loading pattern prior to 30 years of operation.
  • Base load operation, iLe., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule.

DCPP Units 1 and 2 operate as base-load units. PG&E Letter DCL-1 5-1 50 Page 4 of 5 No design changes beyond those identified in general industry guidance or recommended by the original vendors. MRP-227-A states that the recommendations are applicable to all U.S. PWR operating plants as of May 2007 for the three designs considered. There have been no modifications to reactor internals components at DCPP Units 1 or 2 since May 2007. Based on the applicability, as stated for DCPP Units 1 and 2, the criteria, of MRP-227-A, Section 2.4 are met for DCPP Units 1 and 2. In addition to the req*uirement to demonstrate that the criteria of MRP-227-A, Section 2.4 are met, MRP 2013-025 requires that the neutron fluence and heat generation rates for DCPP Units 1 and 2 are within the range of the limiting threshold values defined in MRP 2013-025. The limiting threshold values defined for Westinghouse plants are: Average core power density less than 124 Watts/cm3 Heat generation figure of merit (F) less than or equal to 68 Watts/cm3

  • Active fuel to upper core plate distance greater than 12.2 inches PG&E is currently in the process of evaluating the DCPP Units I and 2 reactor internals components with regard to fuel designs and fuel management according to guidance provided in MRP 2013-025. PG&E is currently scheduled to complete and submit to the NRC the results of this evaluation for DCPP Units 1 and 2 by March 31, 2016 (see Enclosure 4).

References

1. U.S. Nuclear Regulatory Commission Letter, "Summary of January 22-23, 2013, Closed Meeting with the Electric Power Research Institute and Westinghouse," February 21, 2013. (ADAMS: ML13042A048/mlI13043A062).
2. U.S. Nuclear Regulatory Commission Letter, "Summary of February 25, 2013, Telecom with the Electric Power Research Institute and Westinghouse Electric Company," March15, 2013. (ADAMS: ML13067A262).
3. EPRI Letter, MRP 2013-025, "MRP-227-A Applicability Template Guideline,"

October 14, 2013.

4. Materials Reliability Program: Screening, Categorization and Ranking of PWR Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

PG&E Letter DCL-1 5-1 50 Page 5 of 5

5. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175). EPRI, Palo Alto, CA: 2005. 1012081.
6. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA:

2011. 1022863.

7. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internal Components (MRP-232, Rev. 1). EPRI, Palo Alto, CA: 2012. 1021029.

PG&E Letter DCL-1 5-1 50 Regulatory Commitments Pacific Gas and Electric Company (PG&E) is making the following new and revised regulatory commitments (as defined by NEI 99-04) in this submittal: Commitment Due Date PG&E will update the cathodic protection design and installation action plan and associated licensing basis by March 31, 2016 March 31, 2016. PG&E is currently scheduled to complete and submit to March 31, 2016 the NRC an evaluation of the Units I and 2 reactor internals components with regard to fuel designs and fuel management ac~cording to guidance provided in MRP 2013-025.}}

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