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  
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 December 20,2012 PG&E Letter DCL-12-124 Barry S. Allen Site Vice President Diablo Canyon Power Plant Mail Code 104/6 P. O. Box 56 Avila Beach, CA 93424 805.545.4888 Internal: 691.4888 Fax: 805.545.6445 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 1 0 C F R 54.21 (b)

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 December 20,2012 Page 2 PG&E Letter DCL-12-124 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

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.

Affected LRA Section Section 3.4.2.2.2.1 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 Section 3.3.2.1.10 Section 3.3.2.1.11 Table 3.3.1 Table 3.3.2-9 Table 3.3.2-10 Table 3.3.2-11 Table 3.3.2-10 Table 3.4.2-1 PG&E Letter DCL-12-124 Page 2 of2 Reason for Change Updated to incorporate technical and editorial inconsistencies.

Dampers in various heating, ventilation and air conditioning systems have been updated to have the fire barrier intended function be managed by the Fire Protection AMP, 82.1.12, instead of the inspection of internal surfaces in miscellaneous piping and ducting components AMP, 82.1.22.

Updated to address RIS 2012-02. The aging effects and aging 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 PG&E Letter DCL-12-124 Page 2 of 51 Section 3.1 AGING MANAGEMENT OF REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table 3.1.2-2 Reactor Cool Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation -

t SvStl Component Intended Type Function Class 1 Piping PB

<= 4in Class 1 Piping PB

<= 4in Class 1 Piping PB

<= 4in Material Stainless Steel Stainless Steel Stainless Steel Diablo Canyon Power Plant License Renewal Application Environment Aging Effect Requiring Management Borated Water None Lea~age (E~!)

Reactor Coolant Cracking (Int)

Reactor Coolant Loss of material (Int)

Aging Management NUREG-Table 1 Item Notes Program 1801 Vol.

2 Item None IV.E-3 3.1.1.86 A

ASME Section XI IV.C2-1 3.1.1.70 B

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)

Water Chemistry IV.C2-15 3.1.1.83 E, 3 (B2.1.2) and One-Time Inspection (B2.1.16)

PG&E Letter DCL-12-124 Page 3 of 51 Table 3.2.2-1 E ~g meere d Safetv Feat Component Intended Material Type Function Class 1 PB Stainless Piping <= 4in Steel Class 1 PB Stainless Piping <= 4in Steel Class 1 PB Stainless Piping <= 4in Steel Class 1 PB Stainless Piping <= 4in Steel Class 1 PB Stainless Piping <= 4in Steel Valve LBS, PB, Stainless SIA Steel Cast Austenitic Diablo Canyon Power Plant License Renewal Application Environment Borated Water Leakage (Ext)

Reactor Coolant (Int)

Reactor Coolant I (Int)

Treated Borated Water (Int)

Treated Borated Water (Int)

Treated Borated Water (Int)

Section 3.2 AGING MANAGEMENT OF ENGINEERED SAFETY FEATURES S

J f Aaina M,

J t Evaluatl*

Safetv Iniection Svst, Aging Effect Aging Management Program NUREG-Table 1 Notes Requiring 1801 Vol.

Item Management 2 Item None None V.F-13 3.2.1.57 A

Cracking ASME Section Xllnservice IV.C2-1 3.1.1.70 B

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)

Loss of material Water Chemistry (B2.1.2) and IV.C2-15 3.1.1.83 E, 2 One-Time Inspection (B2.1.16)

Loss of material Water Chemistry (B2.1.2) and V.D1-30 3.2.1.49 E, 2 One-Time Inspection (B2.1.16)

Cracking Water Chemistry (B2.1.2) and V.D1-31 3.2.1.48 E, 2 One-Time Inspection (B2.1.16)

Cracking Water Chemistry (B2.1.2) and V.D1-31 3.2.1.48 E, 2 One-Time Inspection (B2.1.16)

PG&E Letter DCL-12-124 Page 4 of 51 Table 3.2.2-4 E ng meere d Safetv Feat, Component Intended Material Type Function

~ bBS, PB, StaiRless s.JA Steel Tubing LSS, PB;- Stainless SIA Steel Valve PS, SIA, Stainless SS Steel Diablo Canyon Power Plant License Renewal Application Environment Glases Gyele GaaliR§ lAfateF

{ffitt Plant Indoor Air (Ext)

Ventilation Atmosphere (Int)

Section 3.2 AGING MANAGEMENT OF ENGINEERED SAFETY FEATURES S

f Aaina M t Evaluatj' Conti' t HVAC Svst, Aging Effect Aging Management Program NUREG-Table 1 Notes Requiring 1801 Vol.

Item Management 2 Item bass af MateFial Glases Gyele GaaliR§ VVateF VII.G~ ~Q d.d.~

. aQ 8

System (B~. ~. ~ Q)

None None VII.J-15 3.3.1.94 A

None None VII.J-15 3.3.1.94 A, 2 PG&E Letter DCL-12-124 Page 5 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS 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 3.3.1.35 I Steel with stainless steel cladding pump casing exposed to treated borated water 3.3.1.47 Steel piping, piping components, piping elements, tanks, and heat exchanger components exposed to closed cycle cooling water Diablo Canyon Power Plant License Renewal Application Loss of material due to cladding breach Loss of material due to general, pitting, and crevice corrosion Recommended A plant-specific agin-9 IYes management program is to be evaluated.

Reference NRC Information Notice 94-63, Boric Acid Corrosion of Charging Pump Casings Caused by Cladding Cracks.

Closed-Cycle Cooling Water System (82.1.10)

No Consistent with t'JUREG 1801.

The plant specific aging management program(s) used to manage the aging include: \\Nater Chemistry (82.1.2) and One Time Inspection (82.1.16).Not 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.

Consistent with NUREG-1801 with aging management program 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 PG&E Letter DCL-12-124 Page 6 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS 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 3.3.1.48 I Steel piping, piping components, piping elements, tanks, and heat exchanger components exposed to closed cycle cooling water 3.3.1.50 I Stainless steel piping, piping components, and piping elements exposed to closed Diablo Canyon Power Plant License Renewal Application Loss of material due to general, pitting, crevice, and galvanic corrosion Loss of material due to pitting and crevice corrosion Closed-Cycle Cooling Water System (B2.1.1 0)

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

Recommended No No abandoned-in-place portions of systems will be managed by Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1.22).

Consistent with NUREG-1801 with aging management program exceptions.

The aging management program(s) with exceptions 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).

Consistent with NUREG-1801 with aging management program exceptions.

PG&E Letter DCL-12-124 Page 7 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS 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 cycle cooling water 3.3.1.72 I Steel HVAC ducting Loss of material due to and components general, pitting, crevice, and internal surfaces (for drip pans and drain lines) exposed to microbiologically influenced condensation (Internal) corrosion Diablo Canyon Power Plant License Renewal Application Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (82.1.22)

Recommended No 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).

Consistent with NUREG-1801 with aging management program exceptions.

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 PG&E Letter DCL-12-124 Page 8 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS 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 3.3.1.91 I Stainless steel and I Loss of material due to pitting I Water Chemistry (82.1.2) steel with stainless

. and crevice corrosion steel cladding piping, piping components, and piping elements exposed to treated borated water Diablo Canyon Power Plant License Renewal Application Recommended No managed by the Fire Protection program 1(82.1.12).

Consistent with NUREG-1801 for material, environment, and aging effect, but a different aging management program Water Chemistry (82.1.2) and One-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).

PG&E Letter DCL-12-124 Page 9 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS 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 PG&E Letter DCL-12-124 Page 10 of 51 3.3.2.1.10 Control Room HVAC System Aging Management Programs Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS 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 PG&E Letter DCL-12-124 Page 11 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS 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 PG&E Letter DCL-12-124 Page 12 of 51 Table 3.3.2-3 Auxif

Svst, oJ J

Component Intended Material Type Function Piping PB Stainless Steel Pump LBS, PB Stainless Steel Pump LBS, PB Stainless Steel Diablo Canyon Power Plant License Renewal Application

~

S

~

~

J f Aaina M oJ -

Environment Aging Effect Requiring Management Plant Indoor Air Loss of material (Int)

Plant Indoor Air None (Ext)

Raw Water (Int)

Loss of material Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS t Evaluatj*

Saltwat, d Chlorination Svst, J

Aging Management NUREG-Table 1 Item Notes Program 1801 Vol.

2 Item Inspection of Internal Wf-3.3.~.2+3.2.1 E Surfaces in 4-V.A-26

.08 Miscellaneous Piping and Ducting Components (B2.1.22)

None VII.J-15 3.3.1.94 A

Open-Cycle Cooling VII.C1-15 3.3.1.79 A

Water System (B2.1.9)

PG&E Letter DCL-12-124 Page 13 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-4 Auxiliary Systems - Summary of Aging Management Evaluation - Component Cooling Water System Component Intended Material Type Function Closure Bolting LBS, PB, Stainless SIA Steel Panel Board SS Carbon Steel Piping LBS, PB, Carbon Steel SIA Sight Gauge PB Glass Tank bBS,PB., Carbon Steel StA

+aRk bBS, PB, GareeR Steel StA

+aRk bBS CareeR Steel Tubing LBS, PB, Stainless SIA Steel Tubing LBS, PB, Stainless SIA Steel Valve bBS,PB., Stainless StA Steel Valve bBS,PB., Stain-less StA Steel Diablo Canyon Power Plant License Renewal Application Environment Atmosphere/

Weather (Ext)

Plant Indoor Air (Ext)

Atmosphere/

Weather (Ext)

Atmosphere/

Weather (Ext)

Closed Cycle Cooling Water (lnt)

PlaRt IREieer Air tE*tj PlaRt IREieer Air

~

Closed Cycle Cooling Water (Int)

Plant Indoor Air (Ext)

Closed Cycle Cooling Water (lnt)

Plant Indoor Air (Ext)

Aging Effect Aging Management NUREG-Table 1 Item Notes Requiring Program 1801 Vol.

Management 2 Item Loss of material Bolting Integrity (B2. 1. 7) None None G

Loss of material External Surfaces VII.I-98 3.3.1.58 B

Monitoring Program (B2.1.20)

Loss of Material External Surfaces VILI-9 3.3.1.58 B

Monitoring Program (B2.1.20)

None None None None AG Loss of material Closed-Cycle Cooling VII.C2-14 3.3.1.47 B

Water System (B2.1.1 0) bess ef material

~~eFRal Sl;lFfases WJ-g d.d.~.a8 g

MeRiteriR§ Pre§ram to') '" ')(\\\\

bess ef material IRspestieR ef IRteFRal VII.G 2d d.d.~.7~

g Sl;lFfases iR MisseliaReel;ls PipiR§ aREi gl;lstiR§ r-to') '" ')')\\

.~-.. --

Loss of material Closed-Cycle Cooling VII.C2-10 3.3.1.50 B

Water System (B2.1.1 0)

None None VII.J-15 3.3.1.94 A

Loss of material Closed-Cycle Cooling VII.C2-10 3.3.1.50 B

Water System (B2.1.1 0)

None None VII.J-15 3.3.1.94 A

PG&E Letter DCL-12-124 Page 14 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-5 Auxiliary Systems - Summary of Aging Management Evaluation - Makeup Water System Component Intended Material Type Function VaNe

.pg Gef:lf:leFAlley Diablo Canyon Power Plant License Renewal Application Environment Aging Effect Requiring Management AtmeSf:lReFet bess ef mateFial V~JeatReF ~E~)

Aging Management NUREG-Table 1 Item Notes Program 1801 Vol.

2 Item E*teFRal SI:JFfases NGAe-Nefle G

MeAiteFiAg PFegFam 10') 1 ')()\\

,~-..

-~

PG&E Letter DCL-12-124 Page 15 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-6 Auxiliary Systems - Summary of Aging Management Evaluation - Nuclear Steam Supply Sampling Svstl Component Intended Material Type Function Valve LBS, PB, ' Stainless SIA Steel Diablo Canyon Power Plant License Renewal Application Environment Plant I ndoor Air (Ext)

Aging Effect Aging Management NUREG-Table 1 Item Notes Requiring Program 1801 Vol.

Management 2 Item None None VII.J-15 3.3.1.94 A

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

Management 2 Item NeRe NeRe VII.J ~a d.d.~.94 A

PG&E Letter DCL-12-124 Page 17 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-8 Auxiliary Systems - Summary of Aging Management Evaluation - Chemical and Volume Control Svstl Component Intended Type Function Class 1 Piping PB

<= 4in Class 1 Piping PB

<= 4in Class 1 Piping PB

<= 4in Vessel LBS Diablo Canyon Power Plant License Renewal Application Material Stainless Steel Stainless Steel Stainless Steel Stainless Steel Environment Aging Effect Requiring Management Borated Water None Leakage (Ext)

Reactor Coolant Loss of material (Int)

Reactor Coolant Cracking (Int)

Treated Borated Loss of material Water (Int)

Aging Management Program NUREG-Table 1 Notes 1801 Vol.

Item 2 Item None VII.J-16 3.3.1.99 A

Water Chemistry (82.1.2) and IV.C2-15 3.1.1.83 E, 5 One-Time Inspection (B2.1.16)

ASME Section Xllnservice IV.C2-1 3.1.1.70 B

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)

Inspection of Internal Surfaces VII.E1-17 3.3.1.91 E, 9 in Miscellaneous Piping and Ducting Components (B2.1.22)}PJateF Gl=lemistpt (B2.1.2) and one Time I

10') '1

'1 a\\

PG&E Letter DCL-12-124 Page 18 of 51 Table 3.3.2-9 Auxif " Svstl Component Intended Material Type Function Damper FB, PB Carbon Steel Damper

~B, PB, Carbon Steel SS Notes for Table 3.3.2-9:

Plant Specific Notes:

S f Aaina M J

Environment Aging Effect Requiring Management Ventilation Loss of material Atmosphere (Int)

Ventilation Loss of material Atmosphere (lnt)

Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS t Evaluaf Miscell HVAC Svstl Aging Management NUREG-Table 1 Item Notes Program 1801 Vol.

2 Item Fire Protection (B2. 1. 12) VII.F4-2 3.3.1.72 E, 3 Inspection of Internal VI I. F4-2 3.3.1.72 8

Surfaces in Miscellaneous Piping and Ducting Components (82.1.22) 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 PG&E Letter DCL-12-124 Page 19 of 51 Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS Table 3.3.2-10 Auxif S

HVAC Svst, Svstl f Aaina M t Evaluat,*

Control R Component Intended Material Environment Aging Effect Aging Management NUREG-Table 1 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 (Galvanized) Atmosphere (Int)

Damper

~P8 Carbon Steel Ventilation Loss of material Inspection of Internal VII.F1-3 3.3.1.72 (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 (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 Notes E, 2, 7 8,2 H,B PG&E Letter DCL-12-124 Page 20 of 51 Table 3.3.2-11 Auxif Component Intended Type Function Damper FB, PB Damper

~PB, SIA, SS Damper FB Damper

~PB, SIA, SS Notes for Table 3.3.2-11:

Plant Specific Notes:

Svstl Material Carbon Steel Carbon Steel Carbon Steel (Galvanized)

Carbon Steel (Galvanized)

S f AGinG M Environment Aging Effect Requiring Management Ventilation Loss of material Atmosphere (lnt)

Ventilation Loss of material Atmosphere (Int)

Ventilation Loss of material Atmosphere (Int)

Ventilation Loss of material Atmosphere (Int)

Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS t Evaluaf Auxif BuildinG HVA C Svstl Aging Management NUREG-Table 1 Notes Program 1801 Vol.

Item 2 Item Fire Protection (B2.1. 12)

VII.F2-3 3.3.1.72 E, 6 Inspection of Internal VII.F2-3 3.3.1.72 B

Surfaces in Miscellaneous Piping and Ducting Components (B2.1.22)

Fire Protection (B2. 1.12)

VII.F2-3 3.3.1.72 E, 2, 6 Inspection of Internal VII.F2-3 3.3.1.72 B,2 Surfaces in Miscellaneous Piping and Ducting Components (B2.1.22) 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 PG&E Letter DCL-12-124 Page 21 of 51 Table 3.3.2-12 Auxif Component Intended Type Function Spray Nozzle SP Spray Nozzle SP Tank PB.,-SS Tank PB.,-SS Valve p.g Diablo Canyon Power Plant License Renewal Application

" Svstl Material Carbon Steel Carbon Steel Carbon Steel Carbon Steel Gast IFeR S

f Aaina M oJ Environment Aging Effect Requiring Management Atmosphere/

Loss of Material Weather (Ext)

Atmosphere/

Loss of Material Weather (Int)

Raw Water (Ext) Loss of material Raw Water (Int)

Loss of material PlaRt IRseeF AiF bess ef FRateFial IfExtt Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS t Evaluatj*

Fire Protection Svstl Aging Management Program NUREG-Table 1 Notes 1801 Vol.

Item 2 Item External Surfaces Monitoring VII.I-8-g 3.3.1.58 B

Program (B2.1.20)

External Surfaces Monitoring VII.I-89 3.3.1.58 B

Program (B2.1.20)

Fire Water System (B2.1.13)

VII.G-24 3.3.1.68 0

Fire Water System (B2.1.13)

VII.G-24 3.3.1.68 D

~*teFRal Sl::lFfases MeRiteFiRQ VUJ..8 a. a. ~.as g

n, IQ') 1 ')(\\\\

~;:"~'

,~-..

-~

I PG&E Letter DCL-12-124 Page 22 of 51 Table 3.3.2-14 Auxif Svst, -

Component Intended Material Type Function

~

pg Gej3j3eF Alley

~

pg Gej3j3eF Alley Valve pg Gej3j3eF Alley Valve LBS, PB Stainless Steel Diablo Canyon Power Plant License Renewal Application Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS S

f Aaina M

- - -------J

--J ----- - ----

- ---- - t Evaluatl' Diesel G tor Svst, Environment Aging Effect Aging Management NUREG-Table 1 Item Notes Requiring Program 1801 Vol.

Management 2 Item QFY Gas ~IRt)

Nooe Nooe

~ d. d.~.9g A

PlaRt IR8eeF AiF Nooe Nooe

~

d.~.~.ad A

,fE*tt 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"'-

,~-.

Plant Indoor Air Loss of material Inspection of Internal VII.F2-1 3.3.1.27 E

(Int)

Surfaces in Miscellaneous Piping and Ducting Components (B2.1.22)

PG&E Letter DCL-12-124 Page 23 of 51 Table 3.3.2-16 Auxif

Svst, Component Intended Material Type Function Valve SIA,-SS Stainless Steel Diablo Canyon Power Plant License Renewal Application S

f Aaina M J.

Environment Aging Effect Requiring Management Dry Gas (Int)

None Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS t Eval G

S Aging Management NUREG-Table 1 Item Notes Program 1801 Vol.

2 Item None VII.J-19 3.3.1.97 A

PG&E Letter DCL-12-124 Page 24 of 51 Table 3.3.2-17 Auxif Svstl J

J Component Intended Material Type Function Valve LBS, PB, Cast Iron SIA (Gray Cast Iron)

Diablo Canyon Power Plant License Renewal Application Section 3.3 AGING MANAGEMENT OF AUXILIARY SYSTEMS S

f Aaina M

t Evaluaf Liauid Rad te Svstl Environment Aging Effect Aging Management NUREG-Table 1 Item Notes Requiring Program 1801 Vol.

Management 2 Item Raw Water (Int)

Loss of material Selective Leaching of VII.C1-11 3.3.1.85 A

Materials (B2.1.17)

PG&E Letter DCL-12-124 Page 25 of 51 Section 3.4

. AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM 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 PG&E Letter DCL-12-124 Page 26 of 51 Section 3.4 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM 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 PG&E Letter DCL-12~124 Page 27 of 51 Section 3.4 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Table 3.4.1 Summary of Aging Management Evaluations in Chapter VIII of NUREG-1801 for Steam and Power Item Component Type Number 3.4.1.04 Steel piping, piping components, and piping elements exposed to treated water Diablo Canyon Power Plant License Renewal Application Aging Effect I Mechanism Aging Management Program Loss of material due to Water Chemistry (82.1.2) and general, pitting and crevice One-Time Inspection corrosion (82.1.16)

Further Discussion Evaluation Recommended Yes Consistent with NUREG-1801 for all non 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 3.4.2.2.2.1.

PG&E Letter DCL-12-124 Page 28 of 51 Section 3.4 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Table 3.4.1 Summary of Aging Management Evaluations in Chapter VIII of NUREG-1801 for Steam and Power Item Component Type Number 3.4.1.06 Steel and stainless steel tanks exposed to treated water Diablo Canyon Power Plant License Renewal Application Aging Effect I Mechanism Aging Management Program Loss of material due to general Water Chemistry (82.1.2) and (steel only) pitting and crevice One-Time Inspection corrosion (82.1.16)

Further Discussion Evaluation Recommended Yes Consistent with NUREG-1801 for all non 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.

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

PG&E Letter DCL-12-124 Page 30 of 51 Section 3.4 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM 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 PG&E Letter DCL-12-124 Page 31 of 51 Section 3.4 AGING MANAGEMENT OF STEAM AND POWER CONVERSION SYSTEM Table 3.4.2-2 Svstl Steam and Power Conversion System - Summary of Aging Management Evaluation - Auxiliary Steam Component Intended Material Type Function Trap LBS Carbon Steel Diablo Canyon Power Plant License Renewal Application Environment Aging Effect Requiring Management Steam-Loss of material fffittSecondary Water (lnt)

Aging Management NUREG-Table 1 Item Notes Program 1801 Vol.

2 Item Inspection of Internal VIII.B1-3.4.1.J+04 E, 8 Surfaces in 8 11 Miscellaneous Piping and Ducting Components (B2.1.22)

PG&E Letter DCL-12-124 Page 32 of 51 Section 3.5 AGING MANAGEMENT OF CONTAINMENTS, STRUCTURES AND COMPONENT SUPPORTS 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 Coatings &

(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 PG&E Letter DCL-12-124 Page 33 of 51 Section 3.6 AGING MANAGEMENT OF ELECTRICAL AND INSTRUMENTATION AND CONTROLS Table 3.6.2-1 Electrical and Instrument and Controls - Summary of Aging Management Evaluation - Electrical ComDonent, Component Intended Material Type Function Metal EC Various Enclosed Bus Metals Used (Bus &

for Electrical Connections)

Contacts Metal EC Various Enclosed Bus Metals Used (Bus &

for Electrical Connections)

Contacts Metal IN Porcelain Enclosed Bus (Insulation &

Insulators)

Metal IN Porcelain Enclosed Bus (Insulation &

Insulators)

Diablo Canyon Power Plant License Renewal Application Environment Atmosphere/

Weather (Ext)

Plant Indoor Air (Ext)

Atmosphere/

Weather (Ext)

Plant Indoor Air (Ext)

Aging Effect Aging Management Program NUREG-Table 1 Notes Requiring 1801 Vol.

Item Management 2 Item Loosening of Aging Management Program for VI.A-11 3.6.1.07 AB bolted Metal Enclosed Bus (B2.1.36) connections Loosening of Aging Management Program for VI.A-11 3.6.1.07 AB bolted Metal Enclosed Bus (B2.1.36) connections Embrittlement, Aging Management Program for VI.A-14 3.6.1.08 AB

cracking, Metal Enclosed Bus (B2.1.36)
melting, discoloration, swelling, or loss of dielectric strength leading to reduced insulation resistance (IR);

electrical failure Embrittlement, Aging Management Program for VI.A-14 3.6.1.08 AB

cracking, Metal Enclosed Bus (B2.1.36)
melting, discoloration, swelling, or loss of dielectric strength leading to reduced insulation resistance (IR);

electrical failure PG&E Letter DCL-12-124 Page 34 of 51 Section 3.6 AGING MANAGEMENT OF ELECTRICAL AND INSTRUMENTATION AND CONTROLS Table 3.6.2-1 Electrical and Instrument and Controls - Summary of Aging Management Evaluation - Electrical ComDonent, Component Intended Material Type Function Metal IN Various Enclosed Bus Insulation (Insulation &

Material Insulators)

(Electrical)

Metal IN Various Enclosed Bus Insulation (Insulation &

Material Insulators)

(Electrical)

Diablo Canyon Power Plant License Renewal Application Environment Atmosphere/

Weather (Ext)

Plant Indoor Air (Ext)

Aging Effect Aging Management Program NUREG-Table 1 Notes Requiring 1801 Vol.

Item Management 2 Item Embrittlement, Aging Management Program for VI.A-14 3.6.1.08 AB

cracking, Metal Enclosed Bus (B2.1.36)
melting, discoloration, swelling, or loss of dielectric strength leading to reduced insulation resistance (IR);

electrical failure Embrittlement, Aging Management Program for VI.A-14 3.6.1.08 AB

cracking, Metal Enclosed Bus (B2.1.36)
melting, discoloration, swelling, or loss of dielectric strength leading to reduced insulation resistance (IR);

electrical failure PG&E Letter DCL-12-124 Page 35 of 51 4.2.1 Neutron Fluence Values Summary Description Section 4 TIME-LIMITED AGING ANALYSES 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 1017 n/cm2 (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 PG&E Letter DCL-12-124 Page 36 of 51 Section 4 TIME-LIMITED AGING ANALYSES 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/cm2 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/cm2 threshold, the nozzles themselves as well as the nozzle to nozzle shell welds remain below the 1 x1 017 n/cm2 threshold throu~h 54 EFPY.

Likewise, the lower shell to lower head weld remains below 1x1 01 n/cm2 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 PG&E Letter DCL-12-124 Page 37 of 51 Section 4 TIME-LIMITED AGING ANALYSES 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 PG&E Letter DCL-12-124 Page 38 of 51 Section 4 TIME-LIMITED AGING ANALYSES 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 PG&E Letter DCL-12-124 Page 39 of 51 Upper Core Plates Section 4 TIME-LIMITED AGING ANALYSES 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 PG&E Letter DCL-12-124 Page 40 of 51 Section 4 TIME-LIMITED AGING ANALYSES 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 PG&E Letter DCL-12-124 Page 41 of 51

4.9 REFERENCES

Section 4 TIME-LIMITED AGING ANALYSES

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 FINAL SAFETY ANALYSIS REPORT SUPPLEMENT PG&E Letter DCL-12-124 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 (Cv 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 1 R+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 PG&E Letter DCL-12-124 Page 43 of 51 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT 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 PG&E Letter DCL-12-124 Page 44 of 51 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT 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 PG&E Letter DCL-12-124 Page 45 of 51 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT 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 tThe 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 PG&E Letter DCL-12-124 Page 46 of 51 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT 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 PG&E Letter DCL-12-124 Page 47 of 51 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Table A4-1 License Renewal Commitments Item # I Commitment 22 I PG&E will:

A. For Reactor Coolant System Nickel-Alloy Pressure Boundary Components:

(1) Implement applicable NRC Orders, Bulletins and Generic Letters associated with nickel-alloys; (2) implement staff-accepted industry guidelines, (3) participate in the industry initiatives, such as owners group programs and the EPRI Materials Reliability Program, for managing aging effects associated with nickel-alloys, and (4) upon completion of these programs, but not less than 24 months before entering the period of extended operation, PG&E will submit an inspection plan for reactor coolant system nickel-alloy pressure boundary components to the NRC for review and approval, and LRA Section 3.1 B. For Reactor Vessel Internals:

I 4.3.3 65 (1) Participate in the industry programs for investigating and managing aging effects on reactor 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 before entering the period of extended operation, PG&E will submit an inspection plan for reactor internals to the NRC for review and approval. PG&E will validate the schedule for inspection of the baffle and former bolts on a plant-specific basis to ensure that it will appropriately manage the design fatigue analysis; :- and (4) in accordance with RIS 2011-07, PG&E will submit Information requested in the safety evaluation for MRP-227 "Pressurized 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.

PG&E will revise the plant procedure on flux thimble tube inspections to reference this letter and 'NeAP 12866 to clarify the technical basis for an adequate margin of safety to ensure that the integrity of the reactor coolant system pressure boundary is maintained. This 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 B2.1.21 Implementation Schedule Concurrent with industry initiatives and upon completion submit an inspection plan and not less than 24 months before entering the period of extended operation.

Information requested in the safety evaluation for MRP-227 will be submitted 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.

Prior to the period of extended ooerationCompleted PG&E Letter DCL-12-124 Page 48 of 51 Appendix A FINAL SAFETY ANALYSIS REPORT SUPPLEMENT Table A4-1 License Renewal Commitments Item #

66 67 68 69 Commitment PG&E will revise its plant procedure to include a 5 percent allowance for predictability and a 10 percent allmvance to account for instrument and wear scar uncertainty. This procedure will also be revised to include an SO percent through wall acceptance criterion based upon its 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 PG&E will update the FSAR in accordance with 10 CFR 50.71 (e) to include the flux thimble tube acceptance criterion. This update is currently scheduled to be included in the next FSAR update, but will be comQLeteg prior to the period of extended operation. Completed PG&E will revise its plant procedure to require the actual plant FTT specific wear data versus wear projections be evaluated every refueling outage to ensure it remains consistent with a maximum non conservative wear projection of 5 percent for wear above 40 percent. If the 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 Marine growth removal and subsequent inspection of all required areas of the Unit 1 and Unit 2 discharge conduits will be completed prior to the period of extended operation. The Unit 2 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 LRA Section B2.1.21 B2.1.21 B2.1.21 82.1.32 Implementation Schedule Prior to the period of extended oDeration Completed Prior to the period of extended

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Completed Prior to the period of extenge9 oDe ration Completed Prior to the period of extended operation PG&E Letter DCL-12-124 Page 49 of 51 82.1.15 Reactor Vessel Surveillance Program Description Appendix 8 AGING MANAGEMENT PROGRAMS 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 (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 1 R4-+23 refueling outage after it has accumulated a fluence equivalent to 9994.2 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 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 I

PG&E letter DCl-12-124 Page 50 of 51 Exceptions to NUREG-1801 None Enhancements None Operating Experience Appendix B AGING MANAGEMENT PROGRAMS 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 (Cv 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 PG&E Letter DCL-12-124 Page 51 of 51 Charpy Upper-Shelf Energy Appendix B AGING MANAGEMENT PROGRAMS 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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