DCL-12-124, 10 CFR 54.21 (B) Annual Update to the Diablo Canyon Power Plant License Renewal Application and License Renewal Application Amendment Number 46

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10 CFR 54.21 (B) Annual Update to the Diablo Canyon Power Plant License Renewal Application and License Renewal Application Amendment Number 46
ML12356A179
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 12/20/2012
From: Allen B
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-12-124
Download: ML12356A179 (55)
v • d • e


Text

Pacific Gas and Electric Company Barry S. Allen Diablo Canyon Power Plant Site Vice President Mail Code 104/6 P. O. Box 56 Avila Beach, CA 93424 805.545 . 4888 December 20,2012 Internal: 691.4888 Fax: 805.545.6445 PG&E Letter DCL-12-124 U.S. Nuclear Regulatory Commission 10 C FR 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 and License Renewal Application Amendment Number 46

Dear Commissioners and Staff:

By letter dated November 23, 2009, Pacific Gas and Electric Company (PG&E) submitted an application to the U.S. Nuclear Regulatory Commission 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 amendment to the LRA must be submitted that identifies any change to the current licensing basis (CLB) that materially affects the contents of the LRA, including the Final Safety Analysis Report Supplement. 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, 2011, through September 30,2012. As a reviewer aid, all pages of the Appendix B aging management program section are provided, including unchanged pages, when there is a change on any of the pages in that section .

PG&E makes no new regulatory commitments (as defined by NEI 99-04) in this letter. Changes to existing commitments are contained in the changes to LRA 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) 545-4160 .

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • San Onofre
  • Wol f Creek

Document Control Desk PG&E Letter DCL-12-124 December 20,2012 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on December 20, 2012.

Sincerely, g a:J ~. 4a----

Barry S. Allen Site Vice President crl/50431321 Enclosures cc: Diablo Distribution cc/enc: Elmo E. Collins, NRC Region IV Regional Administrator Nathanial B. Ferrer, NRC Project Manager, License Renewal Thomas R. Hipschman, NRC Senior Resident Inspector A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • San Onofre
  • Wolf Creek

Enclosure 1 PG&E Letter DCL-12-124 Page 1 of 2 Diablo Canyon Power Plant License Renewal Application (LRA) Changes Reflected in the Annual LRA Update Amendment 46 Affected LRA Section Reason for Change Section 4.2.1 Updated to address the revised date for withdrawal of Unit 1 Section 4.9 Capsule B as accepted in Nuclear Regulatory Commission Section A 1.15 (NRC) letter dated March 2, 2012. (ML120330497)

Section A3.1.1 Section B2.1.15 Section A1.19 Updated to address Pacific Gas and Electric Company (PG&E)

Letter DCL-10-160, dated December 13, 2010. The One-Time Inspection of ASME Code Class 1 Small-Bore Piping aging management program (AMP) will cover piping nominal pipe size less than 4 inches on each Unit.

Section A 1.24 Updated to address PG&E Letter DCL-1 0-73, dated July 7, 2010.

PG&E will inspect all accessible cables, connections, and terminal blocks that are identified with adverse localized environments.

Section A 1.26 Updated to address PG&E Letters DCL-12-148, dated November 24, 201 OJ. and DCL-12-166, dated January 7,2011.

Diablo Canyon Power Plant's (DCPP's) NUREG-1801,Section XI.E3 program includes in-scope inaccessible 480-V and higher power cables. Cable pull boxes with a potential for water intrusion that contain in-scope non-environmental qualification inaccessible 480-V and higher power cables will be inspected and water removal performed at least once every year. Testing of in-scope cables be completed prior to the period of extended operation and will be repeated at least every 6 years thereafter.

Section 4.3.3 Updated to address Regulatory Issue Summary (RIS) 2011-07.

Table A4-1 #22 Information identified in the safety evaluation for MRP-227 "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines", dated June 22,2011, will be submitted to the NRC for review and approval no later than 2 years after issuance of the renewed license and no later than 2 years before the plant enters the period of extended operation, whichever comes first. (ML111600498)

Table A4-1 #65 Completed. The plant procedure on flux thimble tube inspections Table A4-1 #66 has been updated. The DCPP Updated Final Safety Analysis Table A4-1 #67 Report has been updated to include the flux thimble tube Table A4-1 #68 acceptance criterion.

Table A4-1 #69 Completed marine growth removal and subsequent inspection of all required areas of the Unit 1 discharge conduits.

Enclosure 1 PG&E Letter DCL-12-124 Page 2 of2 Affected LRA Section Reason for Change Section 3.4.2.2.2.1 Updated to incorporate technical and editorial inconsistencies.

Section 3.4.2.2.7.1 Table 3.1.2-2 Table 3.2.2-1 Table 3.2.2-4 Table 3.3.1 Table 3.3.2-3 Table 3.3.2-4 Table 3.3.2-5 Table 3.3.2-6 Table 3.3.2-7 Table 3.3.2-8 Table 3.3.2-12 Table 3.3.2-14 Table 3.3.2-16 Table 3.3.2-17 Table 3.4.1 Table 3.4.2-2 Table 3.5.2-3 Table 3.6.2-1 Section 3.3.2.1.9 Dampers in various heating, ventilation and air conditioning Section 3.3.2.1.10 systems have been updated to have the fire barrier intended Section 3.3.2.1.11 function be managed by the Fire Protection AMP, 82.1.12, Table 3.3.1 instead of the inspection of internal surfaces in miscellaneous Table 3.3.2-9 piping and ducting components AMP, 82.1.22.

Table 3.3.2-10 Table 3.3.2-11 Table 3.3.2-10 Updated to address RIS 2012-02. The aging effects and aging Table 3.4.2-1 management programs for three material/environment combinations were changed from "none" to "loss of material, pitting and crevice corrosion" and "Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1.22)",

respectively.

PG&E Letter DCL-12-124 Page 1 of 51 LRA Amendment 46 Affected LRA Sections, Tables, and Figures Table 3.1.2-2 Table 3.2.2-1 Table 3.2.2-4 Table 3.3.1 Section 3.3.2.1.9 Section 3.3.2.1.10 Section 3.3.2.1.11 Table 3.3.2-3 Table 3.3.2-4 Table 3.3.2-5 Table 3.3.2-6 Table 3.3.2-7 Table 3.3.2-8 Table 3.3.2-9 Table 3.3.2-10 Table 3.3.2-11 Table 3.3.2-12 Table 3.3.2-14 Table 3.3.2-16 Table 3.3.2-17 Section 3.4.2.2.2.1 Section 3.4.2.2.7.1 Table 3.4.1 Table 3.4.2-1 Table 3.4.2-2 Table 3.5.2-3 Table 3.6.2-1 Section 4.2.1 Section 4.3.3 Section 4.9 Section A 1.15 Section A 1.19 Section A 1.24 Section A 1.26 Section A3.1.1 Table A4-1 #22,65,66,67, 68,69 Section B2.1.15 Section 3.1 PG&E Letter DCL-12-124 AGING MANAGEMENT OF REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Page 2 of 51 Table 3.1.2-2 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation -

Reactor Cool t SvStl Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Class 1 Piping PB Stainless Borated Water None None IV.E-3 3.1 .1.86 A

<= 4in Steel Lea~age (E~!)

Class 1 Piping PB Stainless Reactor Coolant Cracking ASME Section XI IV.C2-1 3.1 .1.70 B

<= 4in Steel (Int) Inservice Inspection, Subsections IWB, IWC, and IWD for Class 1 components (B2.1.1 )

and Water Chemistry (B2 .1.2) and One-Time Inspection of ASME Code Class 1 Small-Bore Piping (B2.1.19)

Class 1 Piping PB Stainless Reactor Coolant Loss of material Water Chemistry IV.C2-15 3.1.1.83 E, 3

<= 4in Steel (Int) (B2.1.2) and One-Time Inspection (B2.1.16)

Diablo Canyon Power Plant License Renewal Application Section 3.2 PG&E Letter DCL-12-124 AGING MANAGEMENT OF ENGINEERED SAFETY FEATURES Page 3 of 51 Table 3.2.2-1 E ~g meere d Safetv Feat S J f Aaina M,  ::J t Evaluatl* Safetv Iniection Svst, Component Intended Material Environment Aging Effect Aging Management Program NUREG- Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item Class 1 PB Stainless Borated Water None None V.F-13 3.2.1.57 A Piping <= 4in Steel Leakage (Ext)

Class 1 PB Stainless Reactor Coolant Cracking ASME Section Xllnservice IV.C2-1 3.1.1.70 B Piping <= 4in Steel (Int) Inspection, Subsections IWB, IWC, and IWD for Class 1 components (B2.1.1 ) and Water Chemistry (B2.1.2) and One-Time Inspection of ASME Code Class 1 Small-Bore Piping (B2.1.19)

Class 1 PB Stainless Reactor Coolant Loss of material Water Chemistry (B2.1.2) and IV.C2-15 3.1.1.83 E, 2 Piping <= 4in Steel I (Int) One-Time Inspection (B2 .1.16)

Class 1 PB Stainless Treated Borated Loss of material Water Chemistry (B2.1.2) and V.D1-30 3.2.1.49 E, 2 Piping <= 4in Steel Water (Int) One-Time Inspection (B2.1.16)

Class 1 PB Stainless Treated Borated Cracking Water Chemistry (B2.1 .2) and V.D1-31 3.2.1.48 E, 2 Piping <= 4in Steel Water (Int) One-Time Inspection (B2.1 .16)

Valve LBS, PB, Stainless Treated Borated Cracking Water Chemistry (B2 .1.2) and V.D1-31 3.2.1.48 E, 2 SIA Steel Cast Water (Int) One-Time Inspection (B2 .1.16)

Austenitic Diablo Canyon Power Plant License Renewal Application Section 3.2 PG&E Letter DCL-12-124 AGING MANAGEMENT OF ENGINEERED SAFETY FEATURES Page 4 of 51 Table 3.2.2-4 E ng meere d Safetv Feat, S f Aaina M t Evaluatj' Conti' t HVAC Svst, Component Intended Material Environment Aging Effect Aging Management Program NUREG- Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item

~ bBS, PB, StaiRless Glases Gyele bass af MateFial Glases Gyele GaaliR§ VVateF VII.G~ ~Q d.d.~ . aQ 8 s.JA Steel GaaliR§ lAfateF System (B~ . ~ . ~ Q)

{ffitt Tubing LSS, PB;- Stainless Plant Indoor None None VII.J-15 3.3.1.94 A SIA Steel Air (Ext)

Valve PS, SIA, Stainless Ventilation None None VII.J-15 3.3.1 .94 A, 2 SS Steel Atmosphere (Int)

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 5 of 51 Table 3.3.1 Summary of Aging Management Evaluations in Chapter VII of NUREG-1801 for Auxiliary Systems Item I Component Type I Aging Effect I Mechanism I Aging Management Further Discussion Number Program Evaluation Recommended 3.3.1.35 ISteel with stainless Loss of material due to A plant-specific agin-9 IYes Consistent with steel cladding pump cladding breach management program is to be t'JUREG 1801.

casing exposed to evaluated. The plant specific aging treated borated water management program(s)

Reference NRC Information used to manage the aging Notice 94-63, Boric Acid include: \Nater Chemistry Corrosion of Charging Pump (82.1.2) and One Time Casings Caused by Cladding Inspection (82 .1.16) .Not Cracks. applicable. OCPP has replaced all steel with stainless steel clad charging pumps, so the applicable NUREG-1801Iine is no longer used.

See further evaluation in Section 3.3.2.2.14.

3.3.1.47 Steel piping, piping Loss of material due to Closed-Cycle Cooling Water No Consistent with components, piping general, pitting, and crevice System (82.1 .10) NUREG-1801 with aging elements, tanks, and corrosion management program heat exchanger exceptions.

components exposed The aging management to closed cycle cooling program(s) with exceptions water to NUREG-1801 include:

Closed-Cycle Cooling Water System (82.1.10)

A different aging management program is credited for abandoned-in-place components. The aging of internal component surfaces exposed to the closed-cycle cooling water environment of the Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 6 of 51 Table 3.3. 1 Summary of Aging Management Evaluations in Chapter VII of NUREG-1801 for Auxiliary Systems Item I Component Type I Aging Effect I Mechanism I Aging Management Further Discussion Number Program Evaluation Recommended abandoned-in-place portions of systems will be managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1.22) .

3.3.1.48 ISteel piping, piping Loss of material due to Closed-Cycle Cooling Water No Consistent with components, piping general, pitting, crevice, and System (B2 .1.1 0) NUREG-1801 with aging elements, tanks, and galvanic corrosion management program heat exchanger exceptions.

components exposed The aging management to closed cycle cooling program(s) with exceptions water to NUREG-1801 include:

Closed-Cycle Cooling Water System (B2.1.1 0)

A different aging management program is credited for abandoned-in-place components. The aging of internal component surfaces exposed to the closed-cycle cooling water environment of the abandoned-in-place portions of systems will be managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Componf}nts (82.1.22) .

3.3.1.50 I Stainless steel piping, Loss of material due to pitting Closed-Cycle Cooling Water No Consistent with piping components, and crevice corrosion System (B2.1 .10) NUREG-1801 with aging and piping elements management program exposed to closed exceptions.

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 7 of 51 Table 3.3.1 Summary of Aging Management Evaluations in Chapter VII of NUREG-1801 for Auxiliary Systems Item I Component Type I Aging Effect I Mechanism I Aging Management I Further Discussion Number Program Evaluation Recommended cycle cooling water The aging management program(s) with exceptions to NUREG-1801 include:

Closed-Cycle Cooling Water System (82 .1.10)

A different aging management program is credited for abandoned-in-place components. The aging of internal component surfaces exposed to the closed-cycle cooling water environment of the abandoned-in-place portions of systems will be managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1.22) .

3.3.1.72 ISteel HVAC ducting Loss of material due to Inspection of Internal No Consistent with and components general, pitting, crevice, and Surfaces in Miscellaneous NUREG-1801 with aging internal surfaces (for drip pans and drain lines) Piping and Ducting management program exposed to microbiologically influenced Components (82.1.22) exceptions.

condensation (Internal) corrosion The aging management program(s) with exceptions to NUREG-1801 include:

Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1 .22).

Aging management of dampers with a fire barrier intended function are Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 8 of 51 Table 3.3.1 Summary of Aging Management Evaluations in Chapter VII of NUREG-1801 for Auxiliary Systems Item I Component Type I Aging Effect I Mechanism I Aging Management I Further Discussion Number Program Evaluation Recommended managed by the Fire Protection program 1(82. 1. 12).

3.3.1.91 IStainless steel and ILoss of material due to pitting IWater Chemistry (82.1.2) No Consistent with steel with stainless . and crevice corrosion NUREG-1801 for material, steel cladding piping, environment, and aging piping components, effect, but a different aging and piping elements management program Water exposed to treated Chemistry (82.1.2) and One-borated water Time Inspection (82.1.16) is credited.

A different aging management program is credited for abandoned-in-place components. The aging of internal component surfaces exposed to the treated borated water environment of the abandoned-in-place portions of systems will be managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1.22).

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 9 of 51 3.3.2.1.9 Miscellaneous HVAC Systems Aging Management Programs The following aging management programs manage the aging effects for the miscellaneous HVAC systems component types:

  • Fire Protection (82.1 .12)

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 10 of 51 3.3.2.1.10 Control Room HVAC System Aging Management Programs The following aging management programs manage the aging effects for the control room HVAC system component types:

  • Fire Protection (B2. 1. 12)

Diablo Canyon Power Plant License Renewal Application

Enclosure 2 Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 11 of 51 3.3.2.1.11 Auxiliary Building HVAC System "Aging Management Programs The following aging management programs manage the aging effects for the auxiliary building HVAC system component types:

  • Fire Protection (82. 1. 12)

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 12 of 51 Table

- ---- - - 3.3.2-3

- - - -- - Auxif

- - -- - - - - - - oJ Svst,

- J - - - - ~

S~ ~

J f Aaina M oJ -

t Evaluatj* Saltwat, d Chlorination Svst, J

Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes

!

Type Function Requiring Program 1801 Vol.

Management 2 Item Piping PB Stainless Plant Indoor Air Loss of material Inspection of Internal Wf- 3.3.~ .2+3.2. 1 E Steel (Int) Surfaces in 4-V.A-26 .08 Miscellaneous Piping and Ducting Components (B2.1.22)

Pump LBS, PB Stainless Plant Indoor Air None None VII.J-15 3.3.1.94 A Steel (Ext)

Pump LBS, PB Stainless Raw Water (Int) Loss of material Open-Cycle Cooling VII.C1-15 3.3.1.79 A Steel Water System (B2.1.9)

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 13 of 51 Table 3.3.2-4 Auxiliary Systems - Summary of Aging Management Evaluation - Component Cooling Water System Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Closure Bolting LBS, PB, Stainless Atmosphere/ Loss of material Bolting Integrity (B2. 1. 7) None None G SIA Steel Weather (Ext)

Panel Board SS Carbon Steel Plant Indoor Air Loss of material External Surfaces VII.I -98 3.3.1.58 B (Ext) Monitoring Program (B2 .1.20)

Piping LBS, PB, Carbon Steel Atmosphere/ Loss of Material External Surfaces VILI-9 3.3.1.58 B SIA Weather (Ext) Monitoring Program (B2.1.20)

Sight Gauge PB Glass Atmosphere/ None None None None AG Weather (Ext)

Tank bBS,PB., Carbon Steel Closed Cycle Loss of material Closed-Cycle Cooling VII.C2-14 3.3.1.47 B StA Cooling Water (lnt) Water System (B2.1.1 0)

+aRk bBS, PB, GareeR Steel PlaRt IREieer Air bess ef material ~~eFRal Sl;lFfases WJ-g d.d.~ .a8 g StA tE*tj MeRiteriR§ Pre§ram to') '" ')(\\

+aRk bBS CareeR Steel PlaRt IREieer Air bess ef material IRspestieR ef IRteFRal VII.G 2d d.d.~.7~ g

~ Sl;lFfases iR MisseliaReel;ls PipiR§ aREi gl;lstiR§ r- --

....,..., .,......., ,..., ,...., to') . '" . ')')\

.~-

Tubing LBS, PB, Stainless Closed Cycle Loss of material Closed-Cycle Cooling VII.C2-10 3.3.1.50 B SIA Steel Cooling Water (Int) Water System (B2.1.1 0)

Tubing LBS, PB, Stainless Plant Indoor Air None None VII.J-15 3.3.1 .94 A SIA Steel (Ext)

Valve bBS,PB., Stainless Closed Cycle Loss of material Closed-Cycle Cooling VII.C2-10 3.3.1 .50 B StA Steel Cooling Water (lnt) Water System (B2.1.1 0)

Valve bBS,PB., Stain-less Plant Indoor Air None None VII.J-15 3.3.1.94 A StA Steel (Ext)

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 14 of 51 Table 3.3.2-5 Auxiliary Systems - Summary of Aging Management Evaluation - Makeup Water System Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item VaNe .pg Gef:lf:leFAlley AtmeSf:lReFet bess ef mateFial E*teFRal SI:JFfases NGAe- Nefle G V~JeatReF ~E~) MeAiteFiAg PFegFam 10') 1 ')()\

,~- .. -~

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 15 of 51 Table 3.3.2-6 Auxiliary Systems - Summary of Aging Management Evaluation - Nuclear Steam Supply Sampling Svstl Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Valve LBS, PB, ' Stainless Plant Indoor Air None None VII.J-15 3.3.1.94 A SIA Steel (Ext) --

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 16 of 51 Table 3.3.2-7 Auxiliary Systems - Summary of Aging Management Evaluation - Compressed Air System Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item SsleAsis Vall,le pg StaiAless PlaAt IASSSF AiF NeRe NeRe VII.J ~a d.d.~ .94 A Steel fE*tt Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 17 of 51 Table 3.3.2-8 Auxiliary Systems - Summary of Aging Management Evaluation - Chemical and Volume Control Svstl Component Intended Material Environment Aging Effect Aging Management Program NUREG- Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item Class 1 Piping PB Stainless Borated Water None None VII.J-16 3.3.1.99 A

<= 4in Steel Leakage (Ext)

Class 1 Piping PB Stainless Reactor Coolant Loss of material Water Chemistry (82 .1.2) and IV.C2-15 3.1.1.83 E, 5

<= 4in Steel (Int) One-Time Inspection (B2.1.16)

Class 1 Piping PB Stainless Reactor Coolant Cracking ASME Section Xllnservice IV.C2-1 3.1.1 .70 B

<= 4in Steel (Int) Inspection, Subsections IWB, IWC, and IWD for Class 1 components (B2.1.1 ) and Water Chemistry (82.1 .2) and One-Time Inspection of ASME Code Class 1 Small-Bore Piping (B2 .1.19)

Vessel LBS Stainless Treated Borated Loss of material Inspection of Internal Surfaces VII.E1-17 3.3.1.91 E, 9 Steel Water (Int) in Miscellaneous Piping and Ducting Components (B2 .1.22)}PJateF Gl=lemistpt (B2 .1.2) and one Time I 10') '1 '1 a\

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 18 of 51 Table 3.3.2-9 Auxif Svstl S f Aaina M ... t Evaluaf Miscell HVAC Svstl

" " J

"

Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Damper FB, PB Carbon Steel Ventilation Loss of material Fire Protection (B2. 1. 12) VII.F4-2 3.3. 1. 72 E, 3 Atmosphere (Int)

Damper ~B, PB, Carbon Steel Ventilation Loss of material Inspection of Internal VI I. F4-2 3.3.1 .72 8 SS Atmosphere (lnt) Surfaces in Miscellaneous Piping and Ducting Components (82 .1.22)

Notes for Table 3.3.2-9:

Plant Specific Notes:

3 Fire Protection (B2. 1. 12) manages the aging effects associated with this fire damper material and environment combination.

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 19 of 51 Table 3.3.2-10 Auxif Svstl S ,., .., M f Aaina t Evaluat,* Control R HVAC Svst,

" " " "

Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Notes Type Function Requiring Program 1801 Vol. Item Management 2 Item Damper FB, PB Carbon Steel Ventilation Loss of material Fire Protection (B2. 1. 12) VII.F1-3 3.3. 1.72 E, 2, 7 (Galvanized) Atmosphere (Int)

Damper ~P8 Carbon Steel Ventilation Loss of material Inspection of Internal VII.F1-3 3.3.1 .72 8 ,2 (Galvanized) Atmosphere (I nt) Surfaces in Miscellaneous Piping and Ducting Components (82.1.22)

Tubing LBS, SIA Copper Alloy Plant Indoor Air Loss of material Inspection of Internal None None H,B (lnt) Surfaces in Miscellaneous Piping and Ducting Components (B2. 1.22) -- - -

Notes for Table 3.3.2-10:

Plant Specific Notes:

7 Fire Protection (B2.1.12) manages the aging effects associated with this fire damper material and environment combination.

B Interior portions of components exposed to plant indoor air will be conservatively managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program, B2. 1.22 Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 20 of 51 Table 3.3.2-11 Auxif Svstl S f AGinG M t Evaluaf Auxif BuildinG HVA C Svstl Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Notes Type Function Requiring Program 1801 Vol. Item Management 2 Item Damper FB, PB Carbon Steel Ventilation Loss of material Fire Protection (B2.1 . 12) VII.F2-3 3.3. 1.72 E, 6 Atmosphere (lnt)

Damper ~PB, Carbon Steel Ventilation Loss of material Inspection of Internal VII.F2-3 3.3.1.72 B SIA, SS Atmosphere Surfaces in Miscellaneous (Int) Piping and Ducting Components (B2 .1.22)

Damper FB Carbon Steel Ventilation Loss of material Fire Protection (B2. 1.12) VII.F2-3 3.3. 1.72 E, 2, 6 (Galvanized) Atmosphere (Int)

Damper ~PB, Carbon Steel Ventilation Loss of material Inspection of Internal VII.F2-3 3.3.1.72 B,2 SIA, SS (Galvanized) Atmosphere Surfaces in Miscellaneous (Int) Piping and Ducting Components (B2.1.22)

Notes for Table 3.3.2-11:

Plant Specific Notes:

6 Fire Protection (B2. 1. 12) manages the aging effects associated with this fire damper material and environment combination.

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 21 of 51 Table 3.3.2-12

- Auxif Svstl .. S f Aaina M oJ t Evaluatj* Fire Protection Svstl

" " " "

Component Intended Material Environment Aging Effect Aging Management Program NUREG- Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item Spray Nozzle SP Carbon Steel Atmosphere/ Loss of Material External Surfaces Monitoring VII.I-8-g 3.3.1.58 B Weather (Ext) Program (B2.1.20)

Spray Nozzle SP Carbon Steel Atmosphere/ Loss of Material External Surfaces Monitoring VII.I-8 9 3.3.1.58 B Weather (Int) Program (B2.1.20)

Tank PB.,-SS Carbon Steel Raw Water (Ext) Loss of material Fire Water System (B2.1.13) VII.G-24 3.3.1.68 0 Tank PB.,-SS Carbon Steel Raw Water (Int) Loss of material Fire Water System (B2.1 .13) VII.G-24 3.3.1.68 D Valve p.g Gast IFeR PlaRt IRseeF AiF bess ef FRateFial ~*teFRal Sl::lFfases MeRiteFiRQ VUJ..8 a.a. ~ .as g I n,

IfExtt ~;:"~'

IQ') 1 ')(\\

,~- .. -~

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 22 of 51 Table 3.3.2-14 Auxif Svst, - - -- S

- - - - -------J --f Aaina M

- -- --J ----- - ---- - - ---- - t Evaluatl' Diesel G tor Svst, Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item

~ pg Gej3j3eF Alley QFY Gas ~IRt) Nooe Nooe ~ d . d.~ .9g A

~ pg Gej3j3eF Alley PlaRt IR8eeF AiF Nooe Nooe ~ d.~.~ .ad A

,fE*tt Valve pg Gej3j3eF Alley PlaRt IR8eeF AiF Less ef mateFial IRsj3ectieR ef IRtemal WG-9 d . d . ~ .~g fffit1 S~Ffaees iR MisceliaRee~s Pij3iRg aR8 Q~ctiRg

("'. 10') 1 ')')\

-- '1"'- - .. - ,~- . .--

Valve LBS, PB Stainless Plant Indoor Air Loss of material Inspection of Internal VII.F2-1 3.3.1.27 E Steel (Int) Surfaces in Miscellaneous Piping and Ducting Components (B2.1.22)

Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 23 of 51 Table 3.3.2-16

. . . Auxif Svst, S "

f Aaina M J.

t Eval G S Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Valve SIA,-SS Stainless Dry Gas (Int) None None VII.J-19 3.3.1.97 A Steel Diablo Canyon Power Plant License Renewal Application Section 3.3 PG&E Letter DCL-12-124 AGING MANAGEMENT OF AUXILIARY SYSTEMS Page 24 of 51 Table 3.3.2-17 Auxif . --- J Svstl

- J - - - - - - S

- --- - - - - --- --f -Aaina

M

:;,- ----- t Evaluaf Liauid Rad te Svstl Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Valve LBS, PB, Cast Iron Raw Water (Int) Loss of material Selective Leaching of VII.C1-11 3.3.1 .85 A SIA (Gray Cast Materials (B2.1.17)

Iron)

Diablo Canyon Power Plant License Renewal Application Section 3.4 PG&E Letter DCL-12-124 . AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Page 25 of 51 3.4.2.2.2.1 Steel piping and components, tanks, and heat exchangers exposed to treated water and steel piping and components exposed to steam The Water Chemistry program (82.1.2) and the One-Time Inspection program (82.1.16) manages loss of material due to general, pitting, and crevice corrosion for carbon steel and gray cast iron components exposed to secondary water. The one-time inspection includes selected components at susceptible locations where contaminants could accumulate (e.g. stagnant flow locations).

A different aging management program is credited for the main condenser shell and hotwell internal surfaces. The aging of main condenser shell and hotwell internal surfaces exposed to the treated water and steam environment is managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program (82.1.22).

A different aging management program is credited for abandoned-in-place components.

The aging of internal component surfaces exposed to the treated water environment of the abandoned-in-place portions of systems are managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program (82.1.22) .

Diablo Canyon Power Plant License Renewal Application Section 3.4 PG&E Letter DCL-12-124 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Page 26 of 51 3.4.2.2.7.1 Stainless steel, aluminum, and copper alloy piping and components and stainless steel tanks and heat exchangers exposed to treated water The Water Chemistry program (82.1.2) and the One-Time Inspection program (82.1.16) manages loss of material due to pitting and crevice corrosion for stainless steel and copper alloy components exposed to secondary water and demineralized water. The one-time inspection includes selected components at susceptible locations where contaminants could accumulate (e.g. stagnant flow locations).

A different aging management program is *credited for abandoned-in-place piping and components in the auxiliary steam system. The aging of internal component surfaces exposed to the ffiW-treated water environment in the abandoned-in-place portions of the-auxiliary steam systems are managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program (82.1 .22).

Diablo Canyon Power Plant License Renewal Application Section 3.4 PG&E Letter DCL-12~124 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Page 27 of 51 Table 3.4.1 Summary of Aging Management Evaluations in Chapter VIII of NUREG-1801 for Steam and Power Item Component Type Aging Effect I Mechanism Aging Management Further Discussion Number Program Evaluation Recommended 3.4.1.04 Steel piping, piping Loss of material due to Water Chemistry (82.1.2) and Yes Consistent with components, and general, pitting and crevice One-Time Inspection NUREG-1801 for all non piping elements corrosion (82 .1.16) abandoned-in-place exposed to treated components. A different water aging management program is credited for abandoned-in-place components. The aging of internal component surfaces exposed to the treated water environment of the abandoned-in-place portions of systems will be managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82. 1.22).

See further evaluation in Section 3.4.2.2.2.1 .

Diablo Canyon Power Plant License Renewal Application Section 3.4 PG&E Letter DCL-12-124 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Page 28 of 51 Table 3.4.1 Summary of Aging Management Evaluations in Chapter VIII of NUREG-1801 for Steam and Power Item Component Type Aging Effect I Mechanism Aging Management Further Discussion Number Program Evaluation Recommended 3.4.1.06 Steel and stainless Loss of material due to general Water Chemistry (82.1 .2) and Yes Consistent with steel tanks exposed to (steel only) pitting and crevice One-Time Inspection NUREG-1801 for all non treated water corrosion (82 .1.16) abandoned-in-place components. A different aging management program is credited for abandoned-in-place components. The aging of internal component surfaces exposed to the treated water environment of the abandoned-in-place portions of systems will be managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1.22).

See further evaluation in

- - -- ----------- - - -- --- -

Section 3.4.2.2.7. 1.

Diablo Canyon Power Plant License Renewal Application Section 3.4 PG&E Letter DCL-12-124 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Page 29 of-51 Table 3.4.1 Summary of Aging Management Evaluations in Chapter VIII of NUREG-1801 for Steam and Power Item Component Type Aging Effect I Mechanism Aging Management Further Discussion Number Program Evaluation Recommended 3.4.1.24 Steel heat exchanger Loss of material due to Closed-Cycle Cooling Water No Consistent with NUREG-components exposed general, pitting, crevice, and System (82.1 .10) 1801 with aging to closed cycle cooling galvanic corrosion management program water exceptions. **The aging management program(s) with exceptions to NUREG-1801 include: Closed-Cycle Cooling Water System (82.1.10)

A different aging management program is credited for abandoned-in-place components. The aging of internal component surfaces exposed to the closed-cycle cooling water environment of the abandoned-in-place portions of systems will be managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1 .22).

Diablo Canyon Power Plant License Renewal Application Section 3.4 PG&E Letter DCL-12-124 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Page 30 of 51 Table 3.4.2-1 Steam and Power Conversion System - Summary of Aging Management Evaluation - Turbine Steam Suoolv

- - ,- ,- Svst,

- J -

Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Piping LBS Stainless Plant Indoor Air Loss of material Inspection of Internal None None H,6 Steel (lnt) Surfaces in Miscellaneous Piping and Ducting Components (B2. 1.22)

Valve LBS Stainless Plant Indoor Air Loss of material Inspection of Internal None None H,6 Steel (lnt) Surfaces in Miscellaneous Piping and Ducting Components (B2. 1.22)

Notes for Table 3.4.2-1:

Plant Specific Notes:

6 Interior portions of components exposed to plant indoor air will be conservatively managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program, B2. 1.22 Diablo Canyon Power Plant License Renewal Application Section 3.4 PG&E Letter DCL-12-124 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Page 31 of 51 Table 3.4.2-2 Steam and Power Conversion System - Summary of Aging Management Evaluation - Auxiliary Steam Svstl Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Trap LBS Carbon Steel Steam- Loss of material Inspection of Internal VIII.B1- 3.4.1 .J+04 E, 8 fffittSecondary Surfaces in 8 11 Water (lnt) Miscellaneous Piping and Ducting Components (B2.1.22)

Diablo Canyon Power Plant License Renewal Application Section 3.5 PG&E Letter DCL-12-124 AGING MANAGEMENT OF CONTAINMENTS, STRUCTURES AND COMPONENT SUPPORTS Page 32 of 51 Table 3.5.2-3 Containments, Structures, and Component Supports - Summary of Aging Management Evaluation -

Component Intended Material Environment Aging Effect Aging Management NUREG- Table 1 Item Notes Type Function Requiring Program 1801 Vol.

Management 2 Item Fire Barrier FB Fire Barrier - Plant Indoor Air Loss of material Fire Protection (B2.1.12) None None J, 2 Coating s & (Cementitious (Structural) (Ext) cracking Wraps Coating)

Notes for Table 3.5.2-3:

Plant Specific Notes:

2 NUREG-1801 does not provide a line in which Fire Barriers (Ceramic Fiber or Cementitious Coating) are inspected per the Fire Protection program (B2. 1. 12).

Diablo Canyon Power Plant License Renewal Application Section 3.6 PG&E Letter DCL-12-124 AGING MANAGEMENT OF ELECTRICAL AND INSTRUMENTATION AND CONTROLS Page 33 of 51 Table 3.6.2-1 Electrical and Instrument and Controls - Summary of Aging Management Evaluation - Electrical ComDonent, Component Intended Material Environment Aging Effect Aging Management Program NUREG- Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item Metal EC Various Atmosphere/ Loosening of Aging Management Program for VI.A-11 3.6.1.07 AB Enclosed Bus Metals Used Weather (Ext) bolted Metal Enclosed Bus ( B2 .1.36)

(Bus & for Electrical connections Connections) Contacts Metal EC Various Plant Indoor Air Loosening of Aging Management Program for VI.A-11 3.6.1 .07 AB Enclosed Bus Metals Used (Ext) bolted Metal Enclosed Bus (B2 .1.36)

(Bus & for Electrical connections Connections) Contacts Metal IN Porcelain Atmosphere/ Embrittlement, Aging Management Program for VI.A-14 3.6.1.08 AB Enclosed Bus Weather (Ext) cracking, Metal Enclosed Bus (B2.1 .36)

(Insulation & melting, Insulators) discoloration, swelling, or loss of dielectric strength leading to reduced insulation resistance (IR);

electrical failure Metal IN Porcelain Plant Indoor Air Embrittlement, Aging Management Program for VI.A-14 3.6.1.08 AB Enclosed Bus (Ext) cracking, Metal Enclosed Bus (B2 .1.36)

(Insulation & melting, Insulators) discoloration, swelling, or loss of dielectric strength leading to reduced insulation resistance (IR);

electrical failure Diablo Canyon Power Plant License Renewal Application Section 3.6 PG&E Letter DCL-12-124 AGING MANAGEMENT OF ELECTRICAL AND INSTRUMENTATION AND CONTROLS Page 34 of 51 Table 3.6.2-1 Electrical and Instrument and Controls - Summary of Aging Management Evaluation - Electrical ComDonent, Component Intended Material Environment Aging Effect Aging Management Program NUREG- Table 1 Notes Type Function Requiring 1801 Vol. Item Management 2 Item Metal IN Various Atmosphere/ Embrittlement, Aging Management Program for VI.A-14 3.6.1.08 AB Enclosed Bus Insulation Weather (Ext) cracking, Metal Enclosed Bus (B2 .1.36)

(Insulation & Material melting, Insulators) (Electrical) discoloration, swelling, or loss of dielectric strength leading to reduced insulation resistance (IR);

electrical failure Metal IN Various Plant Indoor Air Embrittlement, Aging Management Program for VI.A-14 3.6.1.08 AB Enclosed Bus Insulation (Ext) cracking, Metal Enclosed Bus (B2.1 .36)

(Insulation & Material melting, Insulators) (Electrical) discoloration, swelling, or loss of dielectric strength leading to reduced insulation resistance (IR);

electrical failure - - -- --- - - - - -

Diablo Canyon Power Plant License Renewal Application Section 4 PG&E Letter DCL-12-124 TIME-LIMITED AGING ANALYSES Page 35 of 51 4.2.1 Neutron Fluence Values Summary Description Loss of fracture toughness is an aging effect caused by the neutron embrittlement aging mechanism that results from prolonged exposure to neutron radiation. This process results in increased tensile strength and hardness of the material with reduced toughness. The rate of neutron exposure is defined as neutron flux, and the cumulative degree of exposure over time is defined as neutron fluence. As neutron embrittlement progresses, the toughness/temperature curve shifts down (lower fracture toughness as indicated by Charpy upper-shelf energy or Cv USE), and the curve shifts to the right (brittle/ductile transition temperature increases). Neutron fluence projections are made in order to estimate the effect on these reactor vessel material properties (Section 4.2.2 and Section 4.2.3), and to determine if additional reactor vessel materials will be exposed to fluence greater than 1 x 10 17 n/cm 2 (E>1.0 MeV) as a result of license renewal (extended beltline).

Analysis Unit 1 The last capsule withdrawn and tested from Unit 1 was Capsule V at the end-of-cycle (EOC) 11. At that point, Unit 1 had operated for 14.27 EFPY. This capsule had a lead factor of 2.26 resulting inan exposure equivalent to 32.25 EFPY of operation. The results were documented in WCAP-15958 [Reference 2].

This exposure is less than that expected at EOLE. In PG&E Letter DCL-08-021, PG&E requested a change to the withdrawal date of Unit 1 Capsule B from 20.7 EFPY to 21.9 EFPY in order to capture enough fluence data for EOLE. The change was approved by the NRC in a Safety Evaluation dated September 24, 2008, Diablo Canyon Nuclear Power Plant, Unit No. 1 - Approval of Proposed Reactor Vessel Material Surveillance Capsule Withdrawal Schedule (TAC No. MD8371) [Reference 13].

During the scheduled Unit 1 Sixteenth Refueling Outage (1 R 16), refueling personnel were not able to remove the Capsule B access plug on the reactor core barrel flange.

In PG&E Letter DCL-10-141, dated October 25,2010, PG&E requested a change to the withdrawal date of Unit 1 Capsule B from 21.9 EFPY to 23.2 EFPY. The change was approved by the NRC in a Safety Evaluation dated October 29, 2010, Diablo Canyon Nuclear Power Plant, Unit No. 1 - Approval of Proposed Reactor Vessel Material Surveillance Program Withdrawal Schedule (TAC No. ME4924) [Reference 38].

In PG&E Letter DCL-11-122, dated November 21, 2011 , PG&E requested a change to the withdrawal date of Unit 1 Capsule B from 23.2 EFPY to 33 EFPY to support data acquisition for the EPRI MRP-326, Draft E, "Materials Reliability Progr?m: Coordinated PWR Reactor Vessel Surveillance Program (CRVSP)." The withdrawal date Diablo Canyon Power Plant License Renewal Application

Enclosure 2 Section 4 PG&E Letter DCL-12-124 TIME-LIMITED AGING ANALYSES Page 36 of 51 corresponds to the Unit 1 23rd refueling outage (1 R23), which is scheduled for May 2022. The change was approved by the NRC in a Safety Evaluation dated March 2, 2012, Diablo Canyon Power Plant, Unit No. 1: Safety Evaluation for the Request to Revise the Reactor Vessel Material Surveillance Program Withdrawal Schedule (TAC ME7615) [Reference 41].

Unit 2 The last remaining capsule withdrawn and tested from Unit 2 was Capsule V at EOC 9.

At that point, Unit 2 had operated for 11.49 EFPY. This capsule had a lead factor of

~4.57 resulting in an exposure equivalent to 52.6252.51 EFPY of operation. This exposure is comparable to the predicted EOLE exposure of 54 EFPY, i.e., within the 20 percent limit specified as the acceptance criteria in Regulatory Guide 1.190. The results were documented in WCAP-15423 [Reference 3].

Both Units Based on the guidance specified in Regulatory Guide 1.190, a neutron fluence assessment of the beltline and extended beltline regions was performed by

. Westinghouse in WCAP-17299-NP [Reference 40], for Units 1 and 2, through EOLE.

The peak calculated fast neutron fluence values at the pressure vessel clad/base metal interface are shown in Table 4.2-1 and Table 4.2-2 for Units 1 and 2, respectively.

These fluence data tabulations include fuel cycle specific power distributions through the end of Cycle 16 for Units 1 and 2, as well as fluence projections at several intervals out to 54 EFPY.

The calculations account for a Unit 1 core power uprate from 3338 MWt to 3411 MWt at the onset of Cycle 11. Fluence projections beyond the end of Cycle 16 on Units 1 and 2 are based on the assumption that the spatial core power distributions are defined by the average of Cycles 13-15 for Units 1 and 2 For license renewal, Westinghouse performed additional calculations to define which materials in the DCPP pressure vessels, other than beltline materials, are projected to exceed the threshold neutron fluence of 1x1017 n/cm 2 at 54 EFPY (extended beltline materials). The results of these calculations are documented in WCAP-17299-NP

[Reference 40], for Units 1 and 2, through EOLE For both units, although the nozzle shell course and the associated nozzle shell to intermediate shell weld are projected to exceed the 1x1017 n/cm 2 threshold, the nozzles themselves as well as the nozzle to nozzle shell welds remain below the 1x1 0 17 n/cm 2 threshold throu~h 54 EFPY.

Likewise, the lower shell to lower head weld remains below 1x1 01 n/cm 2 through 54 EFPY for both units.

Table 4.2-3 shows the EOLE fluence values for all beltline and extended beltline materials for both Units 1 and 2.

Diablo Canyon Power Plant License Renewal Application Section 4 PG&E Letter DCL-12-124 TIME-LIMITED AGING ANALYSES Page 37 of 51 As discussed in Section 82.1.15 , both units currently use ex-vessel monitoring dosimetry.

Disposition: Revision, 10 CFR 54.21 (c)(1 )(ii); and Aging Management, 10 CFR 54.21 (c)(1 )(iii)

Revision The fluence projections were revised to quantify expected fluence at the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(ii).

Aging Management Neutron fluence will be monitored and its effects managed for the period of extended operation by the DCPP Reactor Vessel Surveillance program, which is summarized in Section 82.1.15. The validity of these parameters and the analyses that depend upon them will therefore be managed to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

Diablo Canyon Power Plant License Renewal Application Section 4 PG&E Letter DCL-12-124 TIME-LIMITED AGING ANALYSES Page 38 of 51 4.3.3 Fatigue Analyses of the Reactor Pressure Vessel Internals Summary Description The structural adequacy of the reactor internals is discussed in FSAR Section 3.9.3.4.1 .

The reactor internal components are not ASME code components. The reactor internals were designed and built prior to the implementation of Subsection NG of the ASME Boiler and Pressure Vessel Code,Section III, for reactor vessel internals.

Therefore, no plant-specific ASME Code stress report was written during the initial design. However, these components were originally designed to meet the intent of the 1971 Edition of Section III of the ASME Boiler and Pressure Vessel Code with addenda through the Winter 1971. The structural integrity of the reactor internals design has been ensured by analyses performed on both generic and DCPP-specific bases.

Analysis The qualification of the reactor vessel internals was first performed by Westinghouse on a generic basis for 40 years of operation. Some DCPP internal components were subsequently analyzed on a DCPP-specific basis.

Tavg Operating Range Reactor Vessel Internals Analysis In support of the modification to the Tavg operating range, all of the core support structures, except for the upper core plate, lower core plate, and baffle bolts, were qualified based on analyzing the most limiting internal components [Reference 23].

From the four-loop generic stress report, for the applicable components, the most highly stressed due to cyclic thermal loads are:

1. Lower support plate
2. Lower support columns
3. Core barrel nozzles These components therefore have the highest fatigue usage factors and were used to demonstrate compliance of the DCPP reactor internals with the intent of ASME Code,Section III, Subsection NG. The remaining internal components within the scope of the DCPP-specific analysis are bounded by the results of the limiting components and have sufficient margin in the stress and fatigue usage factors to accommodate any expected increases in stress range or number of cycles.

The enhanced DCPP Fatigue Management Program will monitor the 50-year design basis number of transients used in the T avg operating range analysis to ensure it will remain valid for the period of extended operation.

Diablo Canyon Power Plant License Renewal Application Section 4 PG&E Letter DCL-12-124 TIME-LIMITED AGING ANALYSES Page 39 of 51 Upper Core Plates The Unit 2 upper core plate (UCP) was analyzed to support the 2005 Unit 2 upflow conversion modification [Reference 24]. The numbers of transients used in the analysis are bound by the numbers of transients in the current 50-year design basis.

The results of the four-loop generic stress report qualify the Unit 1 UCP for 40 years of operation. However, the results of the DCPP-specific analysis performed for the Unit 2 UCP can be applied to the Unit 1 component, since these components are of similar design [Reference 19].

The enhanced DCPP Fatigue Management Program will monitor the 50-year design basis number of transients used in the Unit 2 upflow conversion modification for the Unit 1 and 2 UCPs to ensure it will remain valid for the period of extended operation.

Lower Core Plates The Unit 1 lower core plate (LCP) was analyzed for the increase in heat generation seen by the lower core plate due to power uprate [Reference 25]. The numbers of transients used in the analysis are bound by the numbers of transients in the current 50-year design basis.

The results of the four-loop generic stress report qualify the Unit 2 LCP for 40 years of operation. However, the results of the DCPP-specific analysis performed for the Unit 1 LCP can be applied to the Unit 2 component, since these components are of similar design [Reference 19].

The enhanced DCPP Fatigue Management Program will monitor the 50-year design basis number of transients used in the Unit 1 power uprate for the Unit 1 and 2 LCPs to ensure it will remain valid for the period of extended operation.

Baffle-Former Bolts The fatigue usage factor of the baffle-former bolts was originally shown to be less than 1.0 based on evaluation of test data which demonstrated acceptable performance for a set of bolt displacements. The adequacy of baffle-former bolts is an industry issue and their extended operation is addressed by participation in industry level initiatives as described below.

Flow Induced Vibration in the Reactor Vessel Internals FSAR Section 3.9.1 and the original SER for DCPP discuss the design and vibration test programs for the reactor vessel internals performed as part of preoperational and startup testing. The dynamic behavior of reactor internals has been studied using experimental data obtained from prototype plants along with results of model tests and static and dynamic tests. Indian Point Nuclear Generating Unit 2 was the prototype for the DCPP Unit 1 internals verification program. Trojan Nuclear Plant data provide additional internals verification for Unit 2 (Unit 1 lower internals are similar to Indian Diablo Canyon Power Plant License Renewal Application Section 4 PG&E Letter DCL-12-124 TIME-LIMITED AGING ANALYSES Page 40 of 51 Point Unit 2; Unit 2 lower internals are similar to Trojan). The tests indicated that no unexpected large vibration amplitudes existed in the internal structure during operation.

The licensing basis does not describe any time limited effects for a licensed operating period associated with flow-induced vibration. Therefore there are no TLAAs, in accordance with 10 CFR 54.3(a) Criteria 2 and 3.

Participation in Industry Programs for Reactor Vessel Internals PG&E will (1) participate in industry programs for inve~tigating and managing the aging effects on the reactor vessel internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; aM (3) upon completion of these programs, but not less than 24 months prior to entering the period of extended operation, PG&E will submit an inspection plan to the NRC for review and approval-:;

and (4) in accordance with RIS 2011-07, PG&E will submit Information requested in the safety evaluation for MRP-227 uPressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," dated June 22, 2011 to the NRC for review and approval no later than 2 years after issuance of the renewed license or no later than 2 years before the plant enters PEG, whichever comes first.

Disposition: Aging Management, 10 CFR 54.21 (c)(1 )(iii)

The design basis number of transients will be managed for the period of extended operation by the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary program, which is summarized in Sections 4.3.1 and B3.1 . Action limits will permit completion of corrective actions before the design basis number of events is exceeded. The continued implementation provides reasonable assurance that fatigue in the reactor vessel internals will be managed for the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(iii).

The integrity of the baffle and former bolts will be managed by the Reactor Vessel Internals Aging Management program, which DCPP committed to implement in LRA Table A4-1, Commitment 22. The implementation of the program provides assurance that fatigue in the baffle and former bolts will be managed for the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(iii).

Diablo Canyon Power Plant License Renewal Application Section 4 PG&E Letter DCL-12-124 TIME-LIMITED AGING ANALYSES Page 41 of 51

4.9 REFERENCES

41. US NRC Letter. Joseph M. Sebrosky, Senior Project Manager, Plant Licensing Branch IV, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation; to Mr. John ' Conway, Senior Vice President - Energy Supply and Chief Nuclear Officer, DCPP. "Diablo Canyon Nuclear Power Plant, Unit No.1:

Safety Evaluation for the Request to Revise the Reactor Vessel Material Surveillance Program Withdrawal Schedule*(TAC No. ME7615}." 2 March 2012.

Diablo Canyon Power Plant License Renewal Application Appendix A PG&E Letter DCL-12-124 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Page 42 of 51 A1.15 Reactor Vessel Surveillance The Reactor Vessel Surveillance program manages loss of fracture toughness due to neutron embrittlement in reactor materials exposed to neutron fluence exceeding 1.0E17 n/cm2 (E>1.0 MeV). The program is consistent with ASTM E 185-70 and ASTM E 185-73 for Units 1 and 2, respectively. Capsules are periodically removed during the course of plant operating life. Neutron embrittlement is evaluated through surveillance capsule testing and evaluation, ex-vessel neutron fluence calculations, and monitoring of reactor vessel neutron fluence. The testing program and reporting conform to requirements of 10 CFR 50 Appendix H, Reactor Vessel Material Surveillance Program Requirements.

Data resulting from the program is used to:

  • Determine pressure-temperature limits, minimum temperature requirements, and end-of-life Charpy upper-shelf energy (C v USE) in accordance with the requirements of 10 CFR 50 Appendix G, Fracture Toughness Requirements; and,
  • Determine end-of-life RT PTS values in accordance with 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock.

The Reactor Vessel Surveillance program provides guidance for removal and testing or storage of material specimen capsules. All capsules that have been withdrawn and tested were stored.

For Unit 1, the last capsule is expected to be withdrawn during the 1R+723 refueling outage after it has accumulated a fluence equivalent to 0094.2 years of operation. The remaining four standby capsules have low lead factors, will remain inside the vessel throughout the vessel lifetime, and will be available for future testing.

There are no capsules remaining in the Unit 2 vessel. All capsules were removed because high lead factors produced exposures comparable to the fluences expected at the end of the period of extended operation.

DCPP Units 1 and 2 currently use ex-vessel monitoring dosimetry, which consists of four gradient chains with activation foils outside the reactor vessel, which will be used to monitor the neutron fluence environment within the beltline region.

Diablo Canyon Power Plant License Renewal Application Appendix A PG&E Letter DCL-12-124 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Page 43 of 51 A1.19 One-Time Inspection of ASME Code Class 1 Small-Bore Piping The One-Time Inspection of ASME Code Class 1 Small-Bore Piping program manages cracking of ASME Code Class 1 piping less than or equal to 4 inches nominal pipe size.

This program is implemented as part of the fourth interval of the DCPP Inservice Inspection (lSI) program.

For ASME Code Class 1 small-bore piping, the lSI program requires volumetric examinations on selected butt weld locations to detect cracking. Weld locations are selected based on the guidelines provided In EPRI TR-112657, Revised Risk-Informed In service Inspection Evaluation Procedure. Volumetric examinations are conducted in accordance with ASME Section XI with acceptance criteria from Paragraph IWB-3000 and IWB-2430. The One-Time Inspection of ASME Code Class 1 Small-Bore Piping program inspections will be completed and evaluated within 10 years prior to the period of extended operation.

Diablo Canyon Power Plant License Renewal Application Appendix A PG&E Letter DCL-12-124 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Page 44 of 51 A1.24 Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements The Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program manages aging effects of electrical cables and connections not subject to 10 CFR 50.49 environmental qualification (EQ) requirements.

Aging effects of embrittlement, melting, cracking, swelling, surface contamination, or discoloration of cables, connections and terminal blocks not subject to 10 CFR 50.49 EQ requirements are evaluated to ensure that cables and connections will continue to perform their intended functions during the period of extended operation.

All cables/cable jackets, connections and terminal blocks within the scope of license renewal in accessible areas with an adverse localized environment are inspected. The inspections of cables, connectors and terminal blocks in accessible areas are representative, within reasonable assurance, of cables, connections and terminal blocks in inaccessible areas within adverse localized environments. At least once every 10 years, cables/cable jackets, connections, and terminal blocks within the scope of license renewal in accessible adverse localized environments are visually inspected for embrittlement, melting, cracking, swelling, surface contamination, or discoloration. The first inspection for license renewal will be completed prior to the period of extended operation.

The acceptance criterion for visual inspection of accessible non-EQ cable jacket, connection and terminal blocks insulating material is the absence of anomalous indications that are signs of degradation. Corrective actions for conditions that are adverse to quality are performed in accordance with the corrective action program as part of the QA program.

The Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program is a new program that will be implemented prior to the period of extended operation. Industry and plant-specific operating experience will be evaluated in the development and implementation of this program.

Diablo Canyon Power Plant License Renewal Application Appendix A PG&E Letter DCL-12-124 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Page 45 of 51 A1.26 Inaccessible Medium Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements The Inaccessible Medium Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements program manages the aging effects of inaccessible medium voltage' 480 volt and higher power cables within the scope of license renewal located in conduit, duct banks, and pull boxes exposed to adverse localized environments caused by significant moisture simultaneously with significant voltage . Significant moisture is defined as periodic exposures to moisture that last more than a few days (e.g., cable in standing water). Periodic exposures to moisture that last less than a few days (i.e., normal rain and drain) are not significant. Significant voltage exposure is defined as being subjected to system voltage for more than twenty five percent of the time.

Cable pull boxes with a potential for water intrusion that contain in-scope non-EQ inaccessible medium voltage 480 volt and higher power cables are inspected for water collection. Collected water is removed as required. This inspection and water removal is performed based on actual plant experience with an inspection frequency of at least once every two yearsyear. Inspection for water collection within the cable pull boxes is performed based on plant experience with water accumulation.

Testing of in scope cables will be performed in accordance with Standard Review Plan for License Renewal, NUREG 1800 Revision 1, Table 3.6 2 which indicates that t The specific type of testing performed on in-scope cables will be determined prior to the initial test, and is to be a proven test for detecting deterioration of the insulation system due to wetting, such as power factor, partial discharge, er-polarization index, .3S-described in EPRI TR 103834 P1 2, or other testing that is state-of-the-art at the time the test is performed. The first test will be completed prior to the period of extended operation and will be repeated every 6 years thereafter.

Diablo Canyon Power Plant License Renewal Application Appendix A PG&E Letter DCL-12-124 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Page 46 of 51 A3.1.1 Neutron Fluence Values Loss of fracture toughness is an aging effect caused by the neutron embrittlement aging mechanism that results from prolonged exposure to neutron radiation. This process results in increased tensile strength and hardness of the material with reduced toughness. The rate of neutron exposure is defined as neutron flux, and the cumulative degree of exposure over time is defined as neutron fluence. As neutron embrittlement progresses, the toughness/temperature curve shifts down (lower fracture toughness as indicated by Charpy upper shelf energy or Cv USE), and the curve shifts to the right (brittle/ductile transition temperature increases). Neutron fluence projections are made in order to estimate the effect on these reactor vessel material properties at the end-of-license extended (EOLE). The basis for EOLE is assumed to be 54 effective full power years (EFPY) based on a lifetime capacity factor of 90 percent for 60 years.

The last capsule withdrawn and tested from Unit 1 was Capsule V at the end-of-cycle (EOC) 11, which yielded an exposure less than that expected at EOLE. Capsule 8 will be withdrawn at ~33 EFPY in order to capture enough fluence data for EOLE. The last remaining capsule withdrawn and tested from Unit 2 was Capsule V at EOC 9, which yielded an exposure comparable to that expected at EOLE.

The fluence values for EOLE were projected using ENDF/8-VI cross sections, and they comply with Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.

The DCPP reactor vessel EOLE fluence projections account for use of lower-leakage cores, and the Unit 1 power uprate. The fluence projections were revised to quantify expected fluence at the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(ii).

Neutron fluence will also be monitored and its effects managed for the period of extended operation by the DCPP Reactor Vessel Surveillance program, described in Section A 1.15. The validity of these parameters and the analyses that depend upon them will therefore be managed to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(iii).

Diablo Canyon Power Plant License Renewal Application Appendix A PG&E Letter DCL-12-124 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Page 47 of 51 Table A4-1 License Renewal Commitments Item # I Commitment LRA Implementation Section Schedule 22 I PG&E will: 3.1 Concurrent with A. For Reactor Coolant System Nickel-Alloy Pressure Boundary Components: industry initiatives (1) Implement applicable NRC Orders, Bulletins and Generic Letters associated with nickel- and upon completion alloys; (2) implement staff-accepted industry guidelines, (3) participate in the industry submit an inspection initiatives, such as owners group programs and the EPRI Materials Reliability Program, for plan and not less managing aging effects associated with nickel-alloys, and (4) upon completion of these than 24 months programs, but not less than 24 months before entering the period of extended operation, before entering the PG&E will submit an inspection plan for reactor coolant system nickel-alloy pressure period of extended boundary components to the NRC for review and approval, and operation .

B. For Reactor Vessel Internals: I 4 .3.3 Information (1) Participate in the industry programs for investigating and managing aging effects on requested in the reactor internals; (2) evaluate and implement the results of the industry programs as safety evaluation for applicable to the reactor internals; aM (3) upon completion of these programs, but not less MRP-227 will be than 24 months before entering the period of extended operation, PG&E will submit an submitted no later inspection plan for reactor internals to the NRC for review and approval. PG&E will validate than 2 years after the schedule for inspection of the baffle and former bolts on a plant-specific basis to ensure issuance of the that it will appropriately manage the design fatigue analysis; :- and (4) in accordance with RIS renewed license or 2011-07, PG&E will submit Information requested in the safety evaluation for MRP-227 no later than 2 years "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," dated before the plant June 22, 2011 to the NRC for review and approval no later than 2 years after issuance of the enters PEG, renewed license or no later than 2 years before the plant enters PEG, whichever comes first. whichever comes first.

65 PG&E will revise the plant procedure on flux thimble tube inspections to reference this letter B2.1 .21 Prior to the period of and 'NeAP 12866 to clarify the technical basis for an adequate margin of safety to ensure extended that the integrity of the reactor coolant system pressure boundary is maintained. This ooeration Completed procedure revision is currently scheduled to be completed prior to December 2011, but will be comoleted orior to the oeriod of extended ooeration ComJ2leted Diablo Canyon Power Plant License Renewal Application Appendix A PG&E Letter DCL-12-124 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Page 48 of 51 Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 66 PG&E will revise its plant procedure to include a 5 percent allowance for predictability and a B2.1 .21 Prior to the period of 10 percent allmvance to account for instrument and wear scar uncertainty. This procedure extended will also be revised to include an SO percent through wall acceptance criterion based upon its oDeration Completed plant specific FTT data wear and NRC acceptance of this SO percent criterion. In conclusion, based on the '-'VCAP 12866 80 percent acceptance criterion, including 5 percent predictability uncertainty and 10 percent for eddy cl:Jrrent testing instrl:Jment and wear scar uncertainty, PG&E will use a net acceptance criterion of 65 percent. This procedure revision is currently scheduled to be completed prior to December 2011, but will be completed prior to the period of extended oDeration. ComJ2/eted 67 PG&E will update the FSAR in accordance with 10 CFR 50.71 (e) to include the flux thimble B2.1.21 Prior to the period of tube acceptance criterion. This update is currently scheduled to be included in the next FSAR extended

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update, but will be comQLeteg prior to the period of extended operation.Completed Completed 68 PG&E will revise its plant procedure to require the actual plant FTT specific wear data versus B2.1 .21 Prior to the period of wear projections be evaluated every refueling outage to ensure it remains consistent with a extenge9 maximum non conservative wear projection of 5 percent for wear above 40 percent. If the oDe ration Completed wear projection for a tube is determined to exceed the 5 p?~cent under pre?iction. and has over 40 percent wear the previous cycle, PG&E will enter It Into the corrective action program for evaluation and disposition . This procedure revision is currently scheduled to be completed prior to December 2011, but will be completed prior to the period of extended oDe ration .Completed 69 Marine growth removal and subsequent inspection of all required areas of the Unit 1 and Unit 82.1 .32 Prior to the period of 2 discharge conduits will be completed prior to the period of extended operation. The Unit 2 extended operation discharge conduit is currently scheduled to be completed during 2R17 (2013). The Unit 1

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Diablo Canyon Power Plant License Renewal Application Appendix 8 PG&E Letter DCL-12-124 AGING MANAGEMENT PROGRAMS Page 49 of 51 82.1.15 Reactor Vessel Surveillance Program Description The Reactor Vessel Surveillance program manages loss of fracture toughness due to neutron embrittlement in reactor materials exposed to neutron fluence exceeding 1.0E17 n/cm 2 (E>1.0 MeV). The program is consistent with ASTM E 185-70 and ASTM E 185-73 for Units 1 and 2, respectively. Capsules are periodically removed during the course of plant operating life. Neutron embrittlement is evaluated through surveillance capsule testing and evaluation, ex-vessel neutron fluence calculations, and monitoring of reactor vessel neutron fluence. The testing program and reporting conform to requirements of 10 CFR 50 Appendix H, Reactor Vessel Material Surveillance Program Requirements. Data resulting from the program is used to:

  • Determine pressure-temperature limits, minimum temperature requirements, and end-of-life Charpy upper-shelf energy (Cv USE) in accordance with the requirements of 10 CFR 50 Appendix G, Fracture Toughness Requirements; and,
  • Determine end-of-life RTPTS values in accordance with 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock.

The Reactor Vessel Surveillance program provides guidance for removal and testing or storage of material specimen capsules. All capsules that have been withdrawn and tested were stored.

For Unit 1, the last capsule is expected to be withdrawn during the 1R4-+23 refueling outage after it has accumulated a fluence equivalent to 9994.2 years of operation. The I remaining four standby capsules have low lead factors, will remain inside the vessel throughout the vessel lifetime, and will be available for future testing.

There are no capsules remaining in the Unit 2 vessel. All capsules were removed because high lead factors produced exposures comparable to the fluence expected at the end of the period of extended operation.

DCPP Units 1 and 2 currently use ex-vessel monitoring dosimetry, which consists of four gradient chains with activation foils outside the reactor vessel, which will be used to monitor the neutron fluence environment within the beltline region.

NUREG-1801 Consistency The Reactor Vessel Surveillance program is an existing program that is consistent with NUREG-1801,Section XI.M31 , Reactor Vessel Surveillance.

Diablo Canyon Power Plant License Renewal Application Appendix B PG&E letter DCl-12-124 AGING MANAGEMENT PROGRAMS Page 50 of 51 Exceptions to NUREG-1801 None Enhancements None Operating Experience Reactor Vessel Surveillance program experience at DCPP is evaluated and monitored to maintain an effective program. This is accomplished by promptly identifying and documenting (using the Corrective Action Program) any conditions or events that could compromise the program. In addition, industry operating experience provides input to ensure that the program is maintained. The DCPP operating experience findings for this program identified no unique plant specific operating experience; therefore DCPP operating experience is consistent with NUREG-1801.

The Reactor Vessel Surveillance program has provided materials data and dosimetry for the monitoring of irradiation embrittlement since plant startup. The use of this program has been reviewed and approved by the NRC during the period of current operation. Surveillance capsules have been withdrawn during the period of current operation, and the data from these surveillance capsules have been used to verify and predict the performance of DCPP reactor vessel beltline materials with respect to neutron embrittlement. Calculations have been performed as required to project the reference temperature for pressurized thermal shock (RT PTS) and Charpy upper-shelf energy (C v USE) values to the end-of-license-extended (EOlE). DCPP pressure-temperature limit curves are valid up to a stated vessel fluence limit, and must be revised prior to operating beyond that limit.

Neutron Fluence The last capsule withdrawn and tested from Unit 1 was Capsule V at the end-of-cycle (EOC) 11, which yielded an exposure less than that expected at EOlE. Capsule B will be withdrawn at ~33 EFPY in order to capture enough fluence data for EOlE. The last capsule withdrawn and tested from Unit 2 was Capsule V at EOC 9, which yielded an exposure comparable to that expected at EOlE. The EOlE fluence projections include the use of lower-leakage cores and the Unit 1 power uprate.

Pressurized Thermal Shock All of the beltline and extended beltline materials in the Diablo Canyon Units 1 and 2 reactor vessels are projected to remain below the PTS screening criteria values of 270°F, for axially oriented welds and plates / forgings, and 300°F, for circumferentially oriented welds (per 10 CFR 50.61), through EOl (32 EFPY) and EOlE (54 EFPY).

Diablo Canyon Power Plant License Renewal Application Appendix B PG&E Letter DCL-12-124 AGING MANAGEMENT PROGRAMS Page 51 of 51 Charpy Upper-Shelf Energy The most recent coupon examination results for both units demonstrate that the DCPP reactor vessel material ages consistently with Regulatory Guide 1.99 predictions and provides a conservative means to satisfy the requirements of 10 CFR 50, Appendix G.

The Cv USE values were revised with projections to the end of the period of extended operation.

The Reactor Vessel Surveillance program operating experience information provides objective evidence to support the conclusion that the effects of aging will be adequately managed so that the component intended functions will be maintained during the period of extended operation.

Conclusion Continued implementation of the Reactor Vessel Surveillance program provides reasonable assurance that the aging effects will be managed so that the systems and components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

Diablo Canyon Power Plant License Renewal Application

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