DCL-10-121, Response to NRC Letter Dated August 25, 2010, Request for Additional Information (Set 19) for License Renewal Application

Response to NRC Letter Dated August 25, 2010, Request for Additional Information (Set 19) for License Renewal Application

ML102700040
Person / Time
Site:	Diablo Canyon
Issue date:	09/22/2010
From:	Becker J Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-10-121
Download: ML102700040 (49)

Text

Pacific Gas and Electric Company James R. Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/5/601 P 0. Box 56 Avila Beach, CA 93424 September 22, 2010 805.545.3462 Internal: 691.3462 PG&E Letter DCL-10-121 Fax: 805.545.6445 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20852 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Letter dated August 25, 2010, Request for Additional Information (Set 19) for the Diablo Canyon License Renewal Application

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 (NRC) for the renewal of Facility Operating Licenses DPR-80 and DPR-82, for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The application included the license renewal application (LRA) and Applicant's Environmental Report - Operating License Renewal Stage.

By letter dated August 25, 2010, the NRC staff requested additional information needed to continue their review of the DCPP LRA.

PG&E's response to the request for additional information is included in Enclosure 1. LRA Amendment 12, resulting from the responses, is included in Enclosure 2 showing the changed pages with line-in/line-out annotations.

PG&E makes no regulatory commitments (as defined in NEI 99-04) in this letter.

If you have any questions regarding this response, please contact Mr. Terence L. Grebel, License Renewal Project Manager, at (805) 545-4160.

I declare under penalty of perjury that the foregoing is true and correct.

James R. Becker A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak

• Diablo Canyon o Palo Verde o San Onofre o South Texas Project o Wolf Creek

Document Control Desk PG&E Letter DCL-10-121 September 22, 2010 Page 2 pns/50338457 Enclosure cc: Diablo Distribution cc/enc: Elmo E. Collins, NRC Region IV Regional Administrator Nathanial Ferrer, NRC Project Manager, License Renewal Kimberly J. Green, NRC Project Manager, License Renewal Michael S. Peck, NRC Senior Resident Inspector Alan B. Wang, NRC Project Manager, Office of Nuclear Reactor Regulation

Enclosure 1 PG&E Letter DCL-10-121 Page 1 of 29 PG&E Response to NRC Letter dated August 25, 2010, Request for Additional Information (Set 19) for the Diablo Canyon License Renewal Application RAI 4.3-1

Background:

In LRA Section 4.3.1, "Cycle Count Action Limits and Corrective Actions" subsection, the applicantidentifies that the corrective actions for the Metal Fatigue of Reactor Coolant PressureBoundary Programif an action limit on the cycle counting of a design basis transient is reached. The applicantstates, in part, that if one of the cycle count action limits is reached, corrective actions will include a review of the fatigue usage calculations will be performed'to ensure that the analyticalbases of the leak-before-break (LBB) fatigue crack propagation analysis is maintained.

The applicant also makes the following statement in LRA Section 4.3.1 to indicate that the action limit on cycle counting would be capable of initiatingcorrective actions in a timely fashion:

"Cycle count action limits have been establishedbased on the design number of cycles. In orderto assure sufficient margin to accommodate occurrence of a low probabilitytransient,,corrective actions must be taken before the remaining number of allowable cycles for any specified transient,including the low-probability, higher-usage-factorevents, becomes less than one. Events which occur more frequently contribute less per event to the usage factor. To account for both cases, corrective actions are required when the cycle count for any of the significant contributors to the usage factor is projected to reach a specified percentage of the design number of cycles before the end of the next fuel cycle."

SRP-LR Section 4.3.2.1.1.3 indicates the program description in GALL AMP X.MI, "MetalFatigue of Reactor Coolant PressureBoundary," states an applicantmay reference the program in GALL AMP X.MI to accept a CUF-basedmetal fatigue TLAA in accordance with the TLAA acceptance criterionin 10 CFR 54.21(c)(1)(iii). The program description in the GALL AMP states that the AMP is an acceptable option for managing metal fatigue for the reactorcoolant pressure boundary (RCPB) components, consideringenvironmental effects. The "scopeof program"element in GALL AMP X.MI states that the scope of the program includes preventive measures to mitigate fatigue cracking of metal components of the reactorcoolant pressure boundary caused by anticipatedcyclic strains in the material.

Issue 1: Fatigue usage calculationsare ASME Section III mandated design calculations.LBB fatigue flaw growth analyses are performedpursuant to the requirement in 10 CFR Part,50, Appendix A, General Design Criterion 4, "Dynamic Effects," and are submitted to the NRC for staff approval. It is not evident how a

Enclosure 1 PG&E Letter DCL-10-121 Page 2 of 29 component fatigue usage factor calculation can be applied to an LBB analysis and how the integrity of the LBB analysis is maintainedby this count.

Request 1: Provide your basis for expanding the, cycle counting activities of the DCPP Metal Fatigueof Reactor Coolant Pressure Boundary AMP to include the 10 CFR 54.2 1(c)(1)(iii) aging management of the LBB TLAA. Identify the design basis transientsaccounted for in the fatigue flaw growth analysis in the LBB. Clarify whether the counting activities will be based on a comparison of the total number of all transientsmonitored for the LBB or on the number of transient types in the LBB. Clarify whether the relationshipbetween the cycle counting activities in the Metal Fatigue of Reactor Coolant Pressure Boundary Programand the LBB is currently accounted for in a plant procedure or in the UFSAR.

Issue 2: The staff notes that, according to the last sentence of the previously quoted material,the applicant will take corrective actions "when the cycle count for any of the significant contributorsto the usage factor is projected to reach a specified percentage of the design number of cycles before the end of the next fuel cycle."

Request 2: Identify all transientsin LRA Table 4.3-2 that are considered to be the significant contributorsto fatigue usage and explain the criteria used to make this determination. Explain why PG&E's cycle count action limit is based on only significant contributorsto fatigue usage and does not account for less significant transients.

Please describe the confirmatory analysis supporting the conclusion that a lower contributingtransientwould not significantlyimpact the CUFs for the components.

PG&E Response to RAI 4.3-1

1. The Diablo Canyon Power Plant (DCPP) Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program (AMP) was expanded to include leak-before-break (LBB) because the LBB fatigue crack growth analysis uses the same type of transients used in the initial design of the nuclear steam supply system, which were used to construct the current DCPP Metal Fatigue of Reactor Coolant Pressure Boundary AMP. J The LBB analysis is described in License Renewal Application (LRA)

Section 4.3.2.12. The transients associated with this analysis are shown below in Table 1 (as provided in DCPP's leak-before-break submittal to NRC dated March 16, 1992, DCL-92-059).

Enclosure 1 PG&E Letter DCL-10-121 Page 3 of 29 The counting activities will be based on a comparison of the number of transient types used in the LBB analysis. The relationship between the cycle counting activities in the Metal Fatigue of Reactor Coolant Pressure Boundary AMP and the LBB is not currently accounted for in a plant procedure or in the Final Safety Analysis Report (FSAR) Update, but it is an enhancement in X.M1 (LRA program B3.1), as stated in LRA Table A4-1.

As shown in DCPP LRA Table 4.3.2, the design transients currently in the DCPP FSAR are:

Table 1 Design Transient 40-yr Transients 50-yr Design, in LBB Analysis Transients Normal RCS heatup and cooldown at *_1 OO 0 F/hr 200 250 Unit loading and unloading at 5 percent 18300 18300 of full power/mmin Step increase and decrease of 10 2000 2500 percent of full power.

Large step load decrease 200 250 Steady state fluctuations 106 Infinite Upset Loss of load (above 15 percent full 80 100 power), without immediate turbine or reactor trip Loss of all offsite power 40 50 Partial loss of flow 80 100 Reactor trip from full power 400 500 Test Conditions Turbine roll test 10 10 Primary side hydrostatic test 5 10 Primary side leak test 50 60 Cold hydrostatic test 10 Not Included

Enclosure 1 PG&E Letter DCL-10-121 Page 4 of 29

2. All transients in LRA Table 4.3-2 are considered to be the significant contributors to fatigue usage and are tracked by the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary AMP (LRA program B3.1), except those transients identified with a "See Note e." Transients which were deemed nonsignificant are those whose stress intensities are low enough to preclude fatigue or those events which are precluded because of DCPP operating practices. These conclusions are supported by the current design or licensing basis analyses (as discussed in LRA Section 4.3.2) and with the use of engineering judgments.

Two transients used in the LBB analysis have been deemed nonsignificant:

(1) Unit loading and unloading at 5 percent of full power/min, and (2) steady state fluctuations. These transients are not counted because consistent with current plant procedures:

a. This transient is associated with load following. The current operating strategy for the DCPP units is continuous base-load power generation. Therefore, the actual number of unit loading/unloading occurrences is expected to be a small fraction of the cycles assumed in the fatigue analyses. Due to the infrequent nature of this cyclic transient, and the large margin to the assumed number of occurrences, it is not necessary to track its occurrence.

b. The number of steady state fluctuation occurrences listed in the FSAR table is "infinite;" therefore, there is no need to count this transient.

Enclosure 1 PG&E Letter DCL-10-121 Page 5 of 29 RAI 4.3-2

Background:

In LRA section 4.3.1, "Cumulative Usage Corrective Actions" subsection, the applicantstates, in part, that if the action limit on the CUF monitoringis reached, corrective actions will include:

1. Determine whether the scope of the Fatigue Management Programmust be enlarged to include additionalaffected reactorcoolant pressure boundary locations. This determination will ensure that other locations do not approach design limits without an appropriateaction.

2. Enhance fatigue managing to confirm continued conformance to the code limit.

Issue 1: Corrective Action (1) is included in the enhancement of LRA Metal Fatigue of Reactor Coolant Pressure Boundary Programin LRA Appendix A Commitment No. 21.

The staff noted that the corrective action is only applicable to reactorcoolant pressure boundary components. However, in its review of LRA Section 4.3.2, the staff confirmed that the TLAA does include the CUF results for some ASME Code Class 2 components that were analyzed to ASME Section III CUF requirements for Code Class 1 components. As a result, the staff noted that the action in CUF monitoring corrective action I may be applicable to the ASME Code Class 2 components analyzed within the scope of the AMP.

Request 1: Verify corrective action.(1) on LRA page 4.3-5, applies to reactorcoolant pressure boundary components, component supports, and ASME Code Class 2 components analyzed to ASME Section III CUF requirementsfor Code Class I components.

Issue 2: Corrective Action 2 of LRA page 4.3-5 states "Enhancefatigue managing to confirm continued conformance to the code limit" Request 2: Clarify what actions would be taken to enhance the fatigue monitoring for this corrective action.

PG&E Response to RAI 4.3-2

1. License Renewal Application (LRA), Appendix A, Commitment Number 21, does not identify ASME Code Class 2 components, as none are included within the scope of the Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program. Specifically, the only ASME Code Class 2 component with an identified time limited aging analysis (TLAA) based on a calculated cumulative use factor (CUF) is the steam generator feedwater nozzles that are discussed in LRA Section 4.3.2.5. These components were replaced in 2008 and 2009 and were analyzed for an additional 50 years of operation.

Enclosure 1 PG&E Letter DCL-10-1.21 Page 6 of 29 Thus, the associated ASME Code Class 1 fatigue analysis is valid through the period of extended operation and the TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(i), not 10 CFR 54.21 (c)(1)(iii).

2. Corrective Action 2 of LRA, page 4.3-5, is not meant to commit to a specific action, but identifies that the methods or assumptions could change (or "be enhanced") to demonstrate that the component is below the ASME Code allowable. For example, the CUF at the location in question: (1) Could be baselined using actual plant historical data in a NB-3200 analysis; (2) The monitoring method could be revised to incorporate revised transients, which removes conservatisms; or (3) Diablo Canyon Power Plant (DCPP) could implement stress-based monitoring, which utilizes six stress tensors or has been appropriately benchmarked. Any corrective actions taken by DCPP to confirm continued conformance with the ASME Code limit will be submitted to the NRC for approval as required.

Enclosure 1 PG&E Letter DCL-10-121 Page 7 of 29 RAI 4.3-3

Background:

LRA Section 4.3.1.1 indicates the applicant will use FatigueProto perform the cycle counting for the applicant'sdesign basis transientsand updates of the CUF values for ASME Section III Code Class 1 components and for those Class 2 components that were conservatively analyzed to ASME Section III CUF requirements for Class I components.

Issue: The staff has confirmed that the use of FatiguePro software is currently accounted for in the applicant's design basis cycle count procedure. The use of FatiguePro applies a one-dimensional Green's function method to compute the stress value inputs for the component CUF values that the software program tracks. The staff addressedpotential non-conservatisms in the ability of FatigueProto perform CUF calculationsin NRC RIS 2008-30, "FatigueAnalysis of Nuclear Power Plant Components," dated December 16, 2008. In RIS 2008-30, the staff recommended that license renewal applicantsperform an analysis to confirm the use of FatigueProwould yield conservative CUF values relative to those that would be generated using the ASME Section III Subarticle NB-3200 methods. The staff notes that the use of FatiguePro is not currently reflected in LRA Commitment No. 21, and the LRA does not provide a basis to determine if the afore mentioned FatiguePro methodology will yield conservative CUF values relative to the use of the methodology describedin ASME Section Il/, Subarticle NB-3200.

Request: Provideyour technical basis to show FatigueProcycle tracking and CUF update methodology generates results more conservative than those generated using the CUF methodology of ASME Section Il/, Subarticle NB-3200. Explain how the Metal Fatigue of Reactor Coolant PressureBoundary Programaddresses the confirmatory analysis, recommended in RIS 2008-30.

PG&E Response to RAI 4.3-3 As described in License Renewal Application (LRA), Section 4.3.1.1 (page 4.3-2), the FatiguePro cycle tracking method (termed cycle based fatigue monitoring method) simply counts the transients to demonstrate the plant is below the analyzed value, thereby demonstrating the code allowable cumulative use factor (CUF) is satisfied.

FatiguePro CUF updates that are credited in the Diablo Canyon Power Plant Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program use cycle based fatigue (CBF) methods, which apply usage to the current CUF based on the actual plant events experienced. The usage accumulated from each event is determined using NB-3200 methods. As stated in LRA, Section 4.3.4 (page 4.3-44, footnote 6), the CBF methods do not use the Green's function, therefore RIS 2008-30 does not apply.

Enclosure 1 PG&E Letter DCL-10-121 Page 8 of 29 RAI 4.3-4

Background:

LRA Section 4.3.1.2 provides the applicant'spresent and projected status of monitored locations. On LRA page 4.3-6, the applicantstates that a "review of the operatinghistory of DCPPUnits 1 and 2 was performed from initial startup to year-end 2008 in orderto baseline the transient event count in the enhanced Fatigue Management Program." In LRA Section 4.3.1.2, BaseliningMethod subsection, (LRA page 4.3-7), the applicantstates that a DCPPspecific procedure defines tracking requirements and recording of plant cyclic transients. The applicantstates that in 1996, FatigueProsoftware was installedat DCPPto monitor and record plant instrumentation in order to identify transientsand that this provided actualplant transient data from the time of the software installation date through 2008, except for a gap in the data from mid-2002 through year-end 2004, which affected the baseline count for the charging and feedwater (FW) cycling transients. LRA Section 4.3.1.2, Baselining Method subsection also provides specific details on the cycle count baselining methods and assumptions for the "AuxiliarySpray during Cooldown" transient,RHR Operation (during Cooldown)" transient,chargingcycling transient,and FW cycling transient.

Issue 1: LRA Section 4.3.1.2 gives no indication about the rigor used to develop the cycle count at DCPP. On page 4.3-7, the applicant only states that "datafrom several sources were considered"for the recount activities.

Request 1: Identify the sources of information used to develop the DCPPtransient operatinghistory.

Issue 2: On LRA page 4.3-7, the applicant states that, after considering the documented sources of cycle counting information, "an explicit cycle count could not be determined for some transients."However, the LRA does not identify which transients are not determined explicitly.

Request 2: Identify the transientsthat were not derived explicitly. Discuss the technical rationale used to derive the 60-year cycle projections for the identified transients.

Issue 3: The applicant'snumber-of-events basis for the 'Auxiliary Spray during Cooldown" transient is given on LRA page 4.3-7. The staff has determined that LRA Table 4.3-2 does not list this transientas within the scope of the design basis transients for this TLAA.

Request 3: Provide the basis for excluding the 'Auxiliary Spray during Cooldown" transientfrom LRA Table 4.3-2.

Issue 4: The applicant'snumber of events basis for the chargingsystem is given at the bottom of LRA page 4.3-7. In LRA Table 4.3-2, the applicant identifies three transients for the chargingsystem (Transients 15, 16, and 17 in the table). The applicantdoes not provide any correlation in the LRA between the number of events basis for charging system on LRA page 4.3-7 and the design basis transientsin LRA Table 4.3-2 that are

Enclosure 1 PG&E Letter DCL-10-121 Page 9 of 29 impacted by this charging system basis. The applicant also.does not specify which quantitative SF was applied to these events orjustify its use in the projection basis.

Request 4: Identify which of the transientsin LRA Table 4.3-2 were assessed in accordance with chargingsystem events basis that was provided at bottom' of page 4.3-

7. Foreach of the transients that were assessed in accordance with this projection basis, identify the SF that was applied to the assessment and justify its use.

PG&E Response to RAI 4.3-4

1. The Diablo Canyon Power Plant (DCPP) transient operating history information was taken from:

• The current plant transient tracking procedure in which transient data is provided by plant operators and is verified by engineering; and

" Computer-assisted cycle counting records (actual plant operating data obtained from the plant process computer).

2. The absence of an event was confirmed for the following transients by interviews with DCPP plant personnel (engineering, operations, and licensing) and by review of reportable events.

" Inadvertent reactor coolant system (RCS) depressurization

• Excessive feedwater flow All other transients which do not include an explicit cycle count are discussed below:

Events Related to Other Counted Events As stated in License Renewal Application (LRA), Section 4.3.1.2 (page 4.3-7), the numbers of events to date for "Auxiliary Spray during Cooldown" and "RHR Operation (during Cooldown)" are based on an assumed number of events per RCS cooldown. Specifically, the "Auxiliary Spray during Cooldown" event generally occurs one or more times late in each cooldown, when normal spray becomes unavailable (because the reactor coolant pumps [RCPs] must be taken off-line at low RCS pressures). It is assumed to occur twice for each counted "Plant Cooldown" event. The "RHR Operation (during Cooldown)" event happens when the RHR system is first brought on-line late in a cooldown (to continue cooling the RCS after the RCPs are stopped). This event is assumed to occur once per "Plant Cooldown" event.

Charging System Events As stated in LRA, Section 4.3.1.2 (page 4.3-7), the numbers of events for the charging system are based on the event frequency for which data is available. The charging system transients include:

Enclosure 1 PG&E Letter DCL-10-121 Page 10 of 29 Transient Number from LRA Table 4.3-2 Transient Description 15 Charging and Letdown, Flow Shutoff and Return to Service (Loop 4 / 3) 16 Loss of Charging with Prompt Return to Service (Loop 4 / 3) 17 Loss of Charging with Delayed Return to Service (Loop 4 / 3) 18 Loss of Letdown with Prompt Return to Service (Loop 4 / 3) 19 Loss of Letdown with Delayed Return to Service (Loop 4 / 3)

FeedwaterCycling As stated in LRA Section, 4.3.1.2 (page 4.3-8), the feedwater cycling events are assumed to correlate to pressurizer heatup cycles. The numbers of events to date was determined by taking the ratio of the number of documented pressurizer heatups through 2008 to the number of expected pressurizer heatups for 60 years of operation and multiplying it by the total number of allowed feedwater cycling events (2,500). For DCPP Unit 1, there were 49 pressurizer heatups through 2008 and 179 total pressurizer heatups projected for 60 years. For DCPP Unit 2, there were 33 pressurizer heatups through 2008 and 179 total pressurizer heatups projected for 60 years.

3. The "Auxiliary Spray during Cooldown" transient discussed on LRA, page 4.3-7, is included in the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program. See revised LRA, Table 4.3-2, in Enclosure 2.

4. The charging system transients are identified in PG&E's response to Request for Additional Information 4.3-4, Part 2, above.

A safety factor of 2.15 is applied in all charging system transient cases to account for the likely higher rate of events during periods for which no actual instrument data is available (prior to FatiguePro installation and from mid-2002 to year-end 2004). It was considered that reactor trips would constitute an extreme example of this effect.

Considering the reactor trips recorded by plant procedures (from 1984 to 2008), the trips during the unmonitored periods occurred 2.15 times more often than during the monitored periods.

Enclosure 1 PG&E Letter DCL-10-121 Page 11 of 29 RAI 4.3-5

Background:

On page LRA 4.3-8, the applicantstates that the projection rate (PR) for the unaccounted periods were performed using both a long-term rate based on the entire transient history for the plant (i.e., number of occurrencessince initialplant startup) and a short term rate for the incrementalcycles that have occurred over the last 10 years. On this page, the applicantstates that the two rates were combined using a weighted average in accordance with the following equation:

PR = [(L TW)*(Iong-term rate) + (STW)*(short-term rate)] / [(L TW) + (STW)],

with L TW being the long-term weighting factor and STW being the short-term weighting factor.

The applicantstates that the values of L TW and STW were determined on an event- or component-specific basis to reflect the most likely future behavior of that event or component.

Issue 1: It is not evident how the L TW and STW values could be derived on a component-specific basis when presumably the design basis CUF calculations for Class I components (andpossibly some Class 2 components analyzed to ASME Section III Class 1 CUF criteria) would involve more than one analyzed transient,and under this basis individual LTW and STW values would have to be assigned to each transientcontributing to the CUF calculation for a given component.

Request 1: Explain the technicalrationale for selection of L TW and STW and how this accommodates events on a component basis.

Issue 2: The PR basis provided on LRA page 4.3-8 only involves a general description about the PR value derivation;the L TW and STW values were derived on a transient-specific (event-specific) or component-specific basis. Thus, the PR basis discussion on LRA page 4.3-8 does not provide the staff with any quantitative basis correlation with the L TW and STW factors used to derive the PRs for the design basis transients in LRA Section 4.3.

Request 2: Identify which transients,in LRA Table 4.3-2, this applies to. Explain how the LTW and STW values were used for the transientprojection basis.

PG&E Response to RAI 4.3-5

1. The long-term weighting (LTW) and short-term weighting (STW) values are not derived on a component-specific basis. The LTW and STW values are only derived on an event-specific basis. See revised License Renewal Application, page 4.3-8, in Enclosure 2.

2. The specific LTW and STW values used for each transient were estimated by taking into account the history of each transient, number of cycles, distribution, and

Enclosure 1 PG&E Letter DCL-10-121 Page 12 of 29 7

qualities of the transient itself. Values were then selected which would likely work.

These values were then compared with the cycle history plot. If the plot showed a projection that fit the past history, the work was done., If not, the weights were adjusted until satisfied with the results.

Assuming no other information, it was assumed that the short-term past was 3 times more likely to predict future performance than the long-term history (i.e., STW = 3, LTW = 1). This was modified based on empirical factors as follows:

• If an event had few total cycles, then the distribution of those events is more likely to reflect random variation than deterministic trends; this would indicate a reduction in the STW relative to the LTW. In cases with very few occurrences, the STW may be reduced to zero - giving a simple linear projection based on the full history.

• If the distribution of past events showed a clear pattern of either increasing or decreasing rate of occurrence, then the STW was increased relative to the LTW.

• STW values were increased for transients relating to planned evolutions (e.g.,

Aux. Spray during CID, RHR Operation and Refueling). Transients that reflect unplanned or accident conditions (e.g., Loss of Power and Loss of Load) had their STW values reduced.

The cycle projections were determined in SIA calculation FP-PGE-305, Cycle and Fatigue Baseline up through YE 2008. Plots of the cycle histories and projections are provided on pages 29-53. A tabulation of the specific STW and LTW values for each transient, taken from that calculation, is reproduced on the following page.

Enclosure 1 PG&E Letter DCL-10-121 Page 13 of 29 DCPP, Unit 1.#cyc as of: 25.00 10.00 Transient 1/01/84 1/01/99 1/01/09 LTR STR LTW STW Rate Aux. Spray during C/D 20 60 78 2.320 1.800 1 3 1.9299 Charging SI into Cold Leg 0 13 15 0.600 0.200 1 2 0.3333 Control Rod Drop 0 0 0 0.000 0.000 1 3 0.0000 High Head SI into CL 0 0 1 0.040 0.100 1 3 0.0850 Inadv. Aux. Spray Actuation 0 2 2 0.080 0.000 .1 0 0.0800 Large Step Load Decrease 0 4 5 0.200 0.100 1 1 0.1500 Loss of All Offsite Power 0 1 1 0.040 0.000 1 1 0.0200 Loss of Load (TT w/o RT) 0 0 5 0.200 0.500 1 1 0.3500 Lp4 Chrg & Ltdn Shutoff 0 2 3 0.120 0.100 1 1 0.1100 Lp4 Chrg Trip Delayed Rtn. 1 3 3 0.080 0.000 1 0 0.0800 Lp4 Chrg Trip Prompt Rtn. 8 48 64 2.240 1.600 1 3 1.7599 Lp4 Ltdn Trip Delayed Rtn. 1 6 6 0.200 0.000 1 0 0.2000 Lp4 Ltdn TripPrompt Rtn. 7 43 60 2.120 1.700 1 3 1.8049 Partial Loss of Flow 0 1 1 0.040 0.000 1 0 0.0400 Plant (RCS) Cooldown 10 30 42 1.280 1.200 1 3 1.2200 Plant (RCS) Heatup 10 31 43 1.320 1.200 1 3 1.2300 Pressurizer Cooldown 10 33 49 1.560 1.600 1 3 1.5900 Pressurizer Heatup 10 34 49 1.560 1.500 1 3 1.5150 RHR Operation (Train A) 10 30 42 1.280 1.200 1 3 1.2200 RHR Operation (Train B) 10 30 42 1.280 1.200 1 3 1.2200 Reactor Trip - C/D no SI 0 10 10 0.400 0.000 1 1 0.2000 Reactor Trip - no C/D 0 45 48 1.920 0.300 1 2 0.8399 Refueling 0 8 14 0.560 0.600 1 3 0.5900 Step Load Decrease 10 percent 0 2 2 0.080 0.000 1 0 0.0800 Step Load Increase 10 percent 0 22 25 1.000 .0.300 2 1 0.7666 Switchover Norm/Alt Charg. 0 0 1 0.040 0.100 1 3 0.0850 Tavg Coastdn to Red.

Temp. . 0 4 6 0.240 0.200 1 3 0.2100

Enclosure 1 PG&E Letter DCL-10-121 Page 14 of 29 DCPP ,Unit 2 I o#2 as of:

.

24.00 1c10.00 Transient 1/01/85d 1/01/99 1/01/09 LTR STR LTW STW Rate Aux. Spray'during C/D 14 42 54 1.667 1.200 1 3 1.3167 Charging SI into Cold Leg 0 13 13 0.542 0.000 1 2 0.1806 Control Rod Drop 0 1 1 0.042 0.000 2 1 0.0278 High Head SI into CL 0 0 0 0.000 0.000 1 3 0.0000 Inadv. Aux. Spray Actuation 4 5 5 0.042 0.000 1 0 0.0417 Large Step Load Decrease 0 3 4 0.167 0.100 1 3 0.1167 Loss of All Offsite Power 0 1 1 0.042 0.000 3 1 0.0312 Loss of Load (TT w/o RT) 0 0 3 0.125 0.300 3 1 0.1687 Lp4 Chrg & Ltdn Shutoff 0 2 4 0.167 0.200 1 3 0.1917 Lp4 Chrg Trip Delayed Rtn. 0 3 3 0.125 0.000 1 1 0.0625 Lp4 Chrg Trip Prompt Rtn. 1 38 54 2.208 1.600 1 3 .1.7521 Lp4 Ltdn Trip Delayed Rtn. 0 4 6 0.250 0.200 1 3 0.2125 Lp4 Ltdn Trip Prompt Rtn. 1 33. 45 1.833 1.200 1 3 1.3583 Partial Loss of Flow 0 2 3 0.125 0.100 1 1 0.1125 Plant (RCS) Cooldown 7 21 30 0.958 0.900 1 3 0.9146 Plant (RCS) Heatup 7 22 31 1.000 0.900 1 3 0.9250 Pressurizer Cooldown 7 21 32 1.042 1.100 1 3 1.0854 Pressurizer Heatup 7 22 33 1.083 1.100 1 3 1.0958 RHR Operation (Train A) 7 21 29 0.917 0.800 1 3 0.8292 RHR Operation (Train B) 7 21 29 0.917 0.800 1 3 0.8292 Reactor Trip - C/D no SI 0 11 11 0.458 0.000 1 2 0.1528 Reactor Trip - no C/D 0 35 37 1.542 0.200 1 2 0.6472 Refueling 0 8 14 "0:583 0.600 1 3 0.5958 Step Load Decrease 10 percent 0 3 5 0.208 0.200 1 3 0.2021 Step Load Increase 10 percnet 0 20 25 1.042 0.500 1 3 0.6354 Switchover Norm/Alt Charg. 0 0 0 0.000 0.000 1 3 0.0000 Tavg Coastdn to Red.

Temp. 0 3 4 0.167 0.100 1 3 0.1167

Enclosure 1 PG&E Letter DCL-10-121 Page 15 of 29 RAI 4.3-6

Background:

LRA Table 4.3-2 provides the'applicant'slist of design basis transients that pertain to the metal fatigue assessments for ASME Code Class 1, 2, or 3 components or components designed to the ANSI B31.1 design specification. UFSAR Table 5.2-4 provides a list of design basis transients for the DCPPunits.

Issue 1: The applicanthas determined that UFSAR Table 4.3-2 provides an accurate -

correlation for all normal operation condition, upset condition, and test condition transients and their design limits in UFSAR Table 5.2-4, with the exception of normal operating condition transient #8, "Tavg Coastdown from Nominal to Reduced Temperature," which currently is not within the scope of LRA Table 4.3-2.

Request 1: Provide your basis for why UFSAR Table 5.2-4, normal operating condition transient #8, 'Tavg Coastdown from Nominal to Reduced Temperature,"is not currently within the scope of LRA Table 4.3-2 and why the applicable 60-year cycle projection data have not been included for this transientin LRA Table 4.3-2.

Issue 2: LRA Table 4.3-2 identifies that the normal operatingcondition transientNos. 5, 13, 14, 15, 16, 17, 18, and 19, and upset condition transient Nos. 24, 26, 27, 28, 29, 30, and 31 are applicable to the scope of the metal fatigue analyses but are not currently within the scope of UFSAR Table 5.2-4.

Request 2: Clarify how these transientsrelate to the scope of the design basis that is currently describedin the DCPP UFSAR (if at all) or applicable design basis procedures or calculations.

Issue 3: LRA Table 4.3-2 includes'transientdata entries for the "DesignBasis Cycles, FSAR Table 5.2-4" and "LimitingAnalyzed Value" columns in the table. The "Limiting Analyzed Value" column is subject to the following Footnote (c) clarification:

"The limiting analyzed value is the lowest number of transients that are considered in DCPPfatigue analyses. The enhanced Fatigue Management Programcompares actual to this limiting analyzed value so that all plant analyses remain valid."

The staff has observed that for those transients in LRA Table 4.3-2 that derive from the list of transientsin UFSAR Table 5.2-4, the value listed in the "LimitingAnalyzed Value" column is sometimes the same as that listed in the "DesignBasis Cycles, FSAR Table 5.2-4" column and sometimes it is lower than the value listed in the "Design Basis Cycles, FSAR Table 5.2-4"column.

Request 3: Clarify which columns (the value in the "DesignBasis Cycles, FSAR Table 5.2-4" column or the value in the "LimitingAnalyzed Value" column) should be relied upon for the design basis transientoccurrence limits.

Enclosure 1 PG&E Letter DCL-10-121 Page 16 of 29 Issue 4: LRA Table 4.3-2 includes test condition transient #37, "Tube Leak Tests." The applicant identifies 800 as the value for the "DesignBasis Cycles, FSAR Table 5.2-4" column and "LimitingAnalyzed Value" column entries for this transient. The staff has determined however, that UFSAR Table 5.2-4 lists this as test condition transient #3.b, and that for this transient,the design basis is broken down into four cases for the transientas follows:

" Case I with a design limit of 400 cycles

" Case 2 with a design limit of 200 cycles

'. Case 3 with a design limit of 120 cycles

• Case 4 with a design limit of 80 cycles Request 4: Justify why the "DesignBasis Cycles, FSAR Table 5.2-4" column and "LimitingAnalyzed Value" column entries in LRA, Table 4.3-2 for "Tube Leak Test" transientare not same as those given in UFSAR Table 5.2-4 for this transient.

Specifically define and discuss each of the Case bases for this transient as defined in UFSAR Table 5.2-4, and explain how DCPParrived at design basis limit values for each'of the Case bases (i.e., for Cases 1 - 4).

PG&E Response to RAI 4.3-6

1. As stated in Diablo Canyon Power Plant (DCPP) Final Safety Analysis Report (FSAR) Update, Revision 18, Table 5.2-4, footnote c, for the replacement steam generators (SGs), "Tavg/power coastdown design transient conditions are enveloped by analyses and evaluations contained in a design change implemented to support operation over a Tavg range of 565 0 F to 577.6 0 F for Cycle 15." As of submittal of this License Renewal Application (LRA), all old SGs had been replaced by replacement SGs. The revised FSAR Update (Revision 19)' was submitted to NRC in 2010 and reflects the removal of this transient from FSAR Table 5.2-4.

Hence, this transient is no longer part of the design basis, does not need to be projected to 60 years, and will not be tracked.

2. Although most of the transients mentioned in Request for Additional Information 4.3-7 are not cited in the FSAR Update, they are used in design basis analyses; and therefore, will conservatively be monitored by the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program (AMP). The basis for inclusion of these transients is shown below:

Enclosure 1 PG&E Letter DCL-10-121 Page 17 of 29 LRA Table 4.3-2 Item Number Transient Name Inclusion in Design Basis Normal 5 Pressurizer Heatup or Identified in the Pressurizer design reports.

Cooldown Cycle in Excess of Tech. Specs. and Within the Bounds of WNEP-8828, Rev. 1 13 Residual Heat Removal Used in the reanalysis of the several Initiation During Cooldown NUREG/CR-6260 locations.

14 Refueling Identified in the analysis Pressurizer surge line for thermal stratification (NRC Bulletin 88-11).

Used in the reanalysis of the several NUREG/CR-6260 locations.

15 Charging and Letdown Flow Used in the re-analysis of the several Shutoff and Return to Service NUREG/CR-6260 locations.

16 Loss of Charging with Prompt Return to Service 17 Loss of Charging with Delayed Return to Service 18 Loss of Letdown with Prompt Return to Service 19 Loss of Letdown with Delayed Return to Service Upset 24 Inadvertent Reactor Coolant Identified in the analysis Pressurizer surge System (RCS) line for thermal stratification (NRC Bulletin Depressurization (Resulting in 88-11).

Reactor Trip)

Used in the reanalysis of the several NUREG/CR-6260 locations.

26 Control Rod Drop Identified in the Pressurizer design reports.

27 Inadvertent Emergency Core Cooling System Actuation Identified in the analysis of the Pressurizer 28 Excessive Feedwater Flow surge line for thermal stratification (NRC Bulletin 88-11).

Used in the reanalysis of the several NUREG/CR-6260 locations.

29 Safety Injection into RCS Cold Used in the reanalysis of the several Leg / High Head Safety NUREG/CR-6260 locations.

Injection 30 Inadvertent Accumulator Blowdown 31 Design Earthquake (OBE) Identified in FSAR Table 5.2-4 as upset I_ . transient #6, Design Earthquake.

Enclosure 1 PG&E Letter DCL-10-121 Page 18 of 29

3. The values in FSAR Table 5.2-4 are the design basis values, meaning all future design work should meet these values. However, this does not mean that all historical fatigue analyses meet these values. During the development of LRA.

Section 4.3, some analyses were identified which were performed using a transient number other than those values in FSAR Table 5.2-4. If the number of transients in the analysis were more limiting than values in FSAR Table 5.2-4, then these values were then incorporated into the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary AMP and were identified in the "Limiting Analyzed Value" column.

The "Limiting Analyzed Value" column should be used when determining what value the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary AMP will count to.

4. The 800 tube leak test cycles, listed in LRA, Table 4.3-2, is the summation of Cases 1 through 4 that are listed in FSAR Table 5.2-4, and was meant to be a simplification for the purposes of the LRA.

FSAR Section 5.2.1.5.5 provides the following details on the "Tube Leak Test" transient.

Case Test Pressure, psig FSAR Table 5.2-4 No.

of Occurences 1 200 400 2 400 200 3 600 120 4 840 80 The current plant cycle counting procedure monitors each of the four cases individually.

Enclosure 1 PG8E Letter DCL-10-121 Page 19 of 29 RAI 4.3-7

Background:

LRA Section 4.3 dispositions the CUF-based TLAAs for many ASME Code Class 1 components by multiplying the CUF values for the components by a factor of 1.2 if the design basis CUF was based on a 50-year design life or by 1.5 if the design basis CUF was based on a 40-year design life. For these TLAAs, DCPPstates that the CUF values remain valid for the period of extended operationin accordance with 10 CFR 54.21(c)(1)(i).

Issue: The multiplication of the design basis CUF by a factor of 1.2 or 1.5 represents a projection of the CUF value for the period of extended operation in that it is changing the CUF value for the component. Thus, components dispositionedin accordance with this methodology should be dispositioned in accordance with the criteriain 10 CFR 54.21(c ) (1) (ii) in that the CUF values have been projected for the period of extended operationand have been found to be acceptable when compared to a CUF value acceptance criterion of 1.0.

Request: Provideyour basis why Class 1 components that are subject to this metal fatigue projection basis have not been dispositionedin accordance with the criterion in 10 CFR 54.21(c)(1)(ii).

PG&E Response to RAI 4.3-7 PG&E agrees that the multiplication of the design basis cumulative use factor (CUF) by a factor of 1.2 or 1.5 represents a projection of the CUF value for the period of extended operation in that it is changing the CUF value for the component. Therefore, the License Renewal Application (LRA) has been revised. See revised LRA Sections 4.1, 4.3, and Appendix A3 in Enclosure 2.

Sections to revise:

• Table 4.1-1 (LRA page 4.1-6 thru 4.1-7) for the below 4.3 sections

• Section 4.3.2.1 (LRA page 4.3-15)

Section 4.3.2.2 (LRA page 4.3-16 & 17)

• Section 4.3.2.3 (LRA page 4.3-20)

Section 4.3.2.4 (LRA page 4.3-24)

• Section 4.3.4 (LRA page 4.3-45)

• LRA Appendix A3 (pages A-29, A-30, A-31, A-34) for the above 4.3' sections

Enclosure 1 PG&E Letter DCL-10-121 Page 20 of 29 RAI 4.3-8

Background:

In the LRA Table 4.3-1, the applicant credits the "Global"monitoring (i.e.

cycle count monitoring) of AMP B3. 1 as the 10 CFR 54.2 1(c)(1)(iii) aging management monitoring basis for dispositioning the CUF analyses for the RPV Core Support Pads, PressurizerSpray Nozzle, and PressurizerHeaterPenetration.

Issue:, LRA Table 4.3-1 or LRA Table 4.3-6 indicatedthat the RPV Core Support Pads, PressurizerSpray Nozzle, and PressurizerHeaterPenetrationin Unit 1 have a maximum limiting design basis CUFs of -0. 89, -0.95, and -0. 94 respectively and limiting 60-year projected CUFs of.- 1.07, -1.14, and -0.9391.

Request: Justify your basis using the "Global"monitoring method of AMP B3. 1 to monitor these components during the period of extended operationin accordance with 10 CFR 54.21(c)(1)(iii), and why it would not be more appropriateto monitor for these components using the CBF monitoring method.

PG&E Response to RAI 4.3-8 A fundamental basis for the Diablo Canyon Power Plant (DCPP) Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program is that as long as the number of transients used in the analysis remain below the analyzed value, then it has be demonstrated that the components are less than the code allowable value and structural integrity is demonstrated.

All transients included in the design basis for the Unit 1 reactor pressure vessel (RPV) core support pads, the pressurizer spray nozzle, and the pressurizer heater penetration are either: (1) counted when the actual transient cycles are experienced by the plant, or (2) determined that the transient used in the design basis does not need to be counted. A list of the transients not being counted, and the basis for not counting them, is included below:

LRA Table 4.3-2 Item Component Transient Number Applicability Name Basis for not Counting the Transient 6 &7 RPV Core Unit loading This transient is associated with load Support Pad and unloading following operation. The current operating at 5 percent of strategy for the DCPP units is continuous Pressurizer full power/min base-load power generation. Therefore, Spray Nozzle the actual number of Unit loading/unloading occurrences is expected to be a small Pressurizer fraction of the cycles assumed in the Heater fatigue analyses. Due to the infrequent Penetration nature of this cyclic transient, and the large margin to the assumed number of occurrences, it is not necessary to track its occurrence.

Enclosure 1 PG&E Letter DCL-10-121 Page 21 of 29 LRA Table 4.3-2 Item Component Transient Number Applicability Name Basis for not Counting the Transient Not RPV Core , Reduced This transient is associated with load Included Support Pad Temperature following operation. DCPP does not Return to operate as a load following plant. Thus, it Pressurizer Power is not necessary to track this transient's Heater occurrence.

Penetration 11 RPV Core Steady State The number of steady state fluctuation Support Pad Fluctuations occurrences listed in the Final Safety Analysis Report table is "infinite;" therefore, Pressurizer there is no need to count this transient.

Heater Penetration Not Pressurizer Boron This transient is associated with load Included Spray Nozzle Concentration following operation. DCPP does not Equalization operate as a load following plant. Thus, it Pressurizer is not necessary to track this transient's Heater occurrence.

Penetration Not Pressurizer Loop Out-of- The loop out-of-service event is not a Included Heater Service credible transient for DCPP because the Penetration DCPP operating licenses do not allow operation with a loop out of service. Thus, this transient does not need to be tracked.

Not Pressurizer Inadvertent This transient is associated with a loop out Included Heater Startup of an of service. The loop out-of-service event is Penetration Inactive loop not a credible transient for DCPP because the DCPP operating licenses do not allow operation with a loop out of service. Thus, this associated transient does not need to be tracked.

Enclosure 1 PG&E Letter DCL-10-121 Page 22 of 29 RAI 4.3-9

Background:

LRA Section 4.3.2.3 (top of pg. 4.3-20) states that the "Unit Loading and Unloading"transient does not need to be counted under the enhanced fatigue management program.

Issue: The staff have determined that the the "UnitLoading and Unloading"transientis within the scope of FSAR Section 5.2.1.5.1 and UFSAR Table 5.2-4, and that Technical Specification (TS) 5.5.5 makes reference to controls to track the FSAR, Section 5.2 and 5.3, cyclic and transientoccurrences to ensure that components are maintained within the design limits. Thus, the staff is of the perception that the counting of this transient would be the activity that correspondsto the control to track the transientunder the TS requirement.

Request: Provideyour basis why the Metal Fatigue of Reactor Coolant Pressure Boundary Program would not need to count this transientduring the period of extended operation when it does appearto be within the scope of the TS tracking requirement.

PG&E Response to RAI 4.3-9 Consistent with current plant procedures, which implement Technical Specification 5.5.5, the current operating strategy for the Diablo Canyon Power Plant (DCPP) units is continuous base-load power generation. Therefore, the actual number of unit loading and unloading occurrences is expected to be a small fraction of the cycles assumed in the fatigue analyses. Due to the infrequent nature of this cyclic transient, and the large margin to the assumed number of occurrences, it is not necessary to track its occurrence.

DCPP cannot change the current operating strategy from continuous base-load power generation to load following since the current design basis does not support load following.

Enclosure 1 PG&E Letter DCL-10-121 Page 23 of 29 RAI 4.3-10

Background:

'LRA Section 4.3.3 provides the fatigue analyses of the reactorpressure vessel internals. The applicantstated that the qualification of reactorvessel internals was first performed by Westinghouse on a generic basis for 40 years of operation. The applicant stated that some DCPPinternalcomponents were subsequently analyzed on a DCPP-specificbasis. The applicantindicated that the lower support plate, lower support columns, and core barrelnozzles had the highest cumulative usage factor values (CUF values) for the reactor vessel internal (RVI components and that the CUFs.

for the remaining RVI components were bounded by the CUF results for these-limiting components. The applicantfurther stated that the enhanced DCPPFatigue Management Program will monitor the 50-year design basis number of transients used in the Tavg operating range analysis to ensure that it remains valid during the period of extended operation.

Issue 1: The staff is unable to determine from the LRA discussion which RVI components were required to be analyzed for fatigue as part of the ASME Section III design.

Request 1: Identify all DCPPRVI components that were required to receive CUF calculationsunder applicableASME Section III design requirements. Forthese components, identify the transientsthat were involved in the calculation of the CUF values and identify what the CUF values are for the components, along with an indication on whether the value for a given RVI component represents an existing design basis value or 60-year projected values. Clarify how the value was calculated if the CUF value for the given RVI components represents a 60-year project value for the TLAA.

Issue 2: The LRA indicated that the fatigue of the RVI components will be managed by the DCPPFatigue Management Programby monitoring the number of transients. The LRA does not provide any justification why it would be acceptable for the applicantto use cycle monitoring of the transients for the lower support plates, lower support columns, and core barrelnozzles as a bounding basis for monitoring the other RVI components that received CUF calculations.

Request 2: Provide your basis for why it is acceptable to use cycle-based monitoring of the transients associatedwith the lower support plates, lower support columns, and core barrelnozzles as a bounding basis for those non-monitored RVI components with CUF values.

Enclosure 1 PG&E Letter DCL-10-121 Page 24 of 29 PG&E Response to RAI 4.3-10

1. The reactor vessel internals components presented in the table below are those that were required to receive cumulative use factor (CUF) calculations applicable to ASME Section III design requirements. The table also lists the results of the Diablo Canyon Power Plant (DCPP)-specific fatigue analyses'for the existing plant design basis (i.e., 50 years).

Component 50-Year Usage Factor for 50-Year Usage Factor for Unit 1 (existing design basis) Unit 2 (existing design basis)

Lower Support Plate - 0.52 0.706 Atypical Region Lower Support 0.945 0.486 Columns Core Barrel Nozzle-Section A-A 0.413 0.413 Lower Support 0.388 0.388 Lower Core Plate 0.52 0.52, Upper Core Plate 0.8 0.88 Baffle Bolts <1.0 _<1.0 The table below presents the transients that were used to calculate the CUF values above.

Design Transients Normal Design Number of Cycles Plant Heatup .250 Plant Cooldown 250 Unit Loading at 5 Percent/Min 18,300 Unit Unloading at 5 Percent/Min 18,300 Step Load Increase of 10 Percent of Full Load 2500 Step Load Decrease of 10 Percent of Full Load 2500 Large Step Decrease with Steam Dump 250

Enclosure 1 PG&E Letter DCL-10-121 Page 25 of 29 Design Transients Upset

.Loss of Load w/o Immediate Turbine or Reactor Trip 100 Loss of Power 50 Partial Loss of Flow 100 Reactor Trip from Full Power 500 Design Earthquake 20

2. A fundamental basis for the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary Aging Management Program is that as long as the number of transients used in the analysis remain below the analyzed value, then it has been demonstrated that the components are less than the code allowable value, and structural integrity is demonstrated.

All transients included in the design basis for the lower support plates, lower support columns, and core barrel nozzles are either: (1) counted when the actual transient cycle is experienced by the plant, or (2) determined that the transient used in the design basis does not need to be counted. A list of the transients used in the reactor vessel internals analyses that are not being counted and the basis for not counting them is included below:

LRA Table 4.3-2 Item Transient Number Name Basis for not Counting the Transient 6 &7 Unit loading This transient is associated with load following operation.

and unloading The current operating strategy for the DCPP units is at 5 percent of continuous base-load power generation. Therefore, the full power/min actual number of unit loading/unloading occurrences is expected to be a small fraction of the cycles assumed in the fatigue analyses. Due to the infrequent nature of this cyclic transient, and the large margin to the assumed number of occurrences, it is not necessary to track its occurrence.

d

-, Enclosure 1 PG&E Letter DCL-10-121 Page 26 of 29 RAI 4.3-11

Background:

The GALL Report states that the AMP addresses the effects of coolant environment by applying an environmental life correction factors to existing ASME code fatigue analyses based on factors.in NUREG/CR-6583 and NUREG/CR-5704, or appropriatealternativemethods.

Issue: The applicanthas stated that the environmental factors are determined by NUREG/CR-6583 and NURGE/CR-5704, or appropriatealternative methods.

Request: Clarify what appropriatealternative method would be used to calculate the-environmental factors for fatigue calculations.

PG&E Response to RAI 4.3-11 This statement regarding "appropriate alternative methods" is not meant to commit to a specific method, but merely identifies that alternative methods exist (as stated in NUREG-1 801) to calculate environmentally assisted fatigue factors (Fen). As stated in License Renewal Application, Section 4.3.4, to address the environmental effects on fatigue, Diablo Canyon Power Plant (DCPP) used material-specific guidance presented in NUREG/CR-6583 and NUREG/CR-5704.

The determination of an "appropriate alternative method" can only be made by the NRC. Therefore, if DCPP uses an "appropriate alternative method" in the future, it would require the approval of the NRC.

Enclosure 1 PG&E Letter DCL-10-121 Page 27 of 29 RAI 4.3-12

Background:

10 CFR Part 54.21 states that each application must identify and list those structures and components subject to an aging management review.

Issue 1: LRA Section 4.3 indicates that the following components were required to be analyzed in accordance with an applicable CUF analysis; however, the AMR Tables in LRA Section 3.1 do not appearto include applicable AMR items that address cumulative fatigue damage for the components:

• RV core support lugs orpads (as indicated in LRA Table 4.3-1)

• RV inlet and outlet nozzle support pads (as indicated in LRA Table 4.3-1

• Reactor coolant pump (RCP)casings (as indicated in LRA Section 4.3.2.3)

" Valve support bracket for the Unit 2 pressurizer(as indicatedin LRA Table 4.3-6)

" SG primary manway, secondary, and feedring components (as indicated in LRA Table 4.3-7)

• RV internallower support plate, lower support columns, core barrelnozzles, and baffle-former plates (as indicated in LRA Section 4.3.3)

Request 1: Provideyour basis why the AMR tables in LRA Section 3.1 do not appearto include any AMR items addressing the management of cumulative fatigue damage for these components.

Issue 2: The staff have noted that the LRA includes AMRs on cumulative fatigue damage only for ASME Section III Class 2 or 3 or ANSI B31.1 piping in the following balance of plant emergency safety feature (ESF), auxiliarysystem (AUX), and steam and power conversion subsystems:

" Safety Injection System (LRA Table 3.2.2-1)

" RHR System (LRA Table 3.2.2-3)

• Chemical and Volume Control System (LRA Table 3.3.2-8)

• Turbine Steam Supply System (LRA Table 3.4.2-1)

• FeedwaterSystem (LRA Table 3.4.2-3)

• Auxiliary FeedwaterSystem (LRA Table 3.4.2-5)

Request 2: Provide your basis why the AMR tables in LRA Sections 3.2, 3.3, and 3.4 do not appearto include any AMR items addressingcumulative fatigue damage for the ANSI B31. 1 or B31.7 piping components in the systems:

" LRA Table 3.2.2-2, Containment Spray System AMRs

" LRA Table 3.2.2-4, Containment HVAC System AMRs

• All Table 2 AMR Tables forAUX subsystems in LRA Section 3.2 otherthan that for Table 3.3.2-8, Chemical and Volume Control System

" LRA Table 3.4.2-2, Auxiliary Steam System

" LRA Table 3.4.2-4, Condensate System

Enclosure 1 PG&E Letter DCL-10-121 Page 28 of 29 PG&E Response to RAI 4.3-12

1. See revised License Renewal Application (LRA), Tables 3.1.2-1 and 3.1.2-4, in Enclosure 2 to add aging management review (AMR) items for the following components subject to cumulative fatigue damage:

" Reactor vessel (RV) nozzle support pads

" Reactor vessel and ihternals (RVI) core barrel assembly

• RV core support lugs

• Valve support bracket (Unit 2 only)

" Steam generator (SG) secondary manway and handhole covers

• SG feedwater ring

• SG primary manway covers The following list specifies those components that did not require LRA revisions and the logic for this determination:

Reactor coolant pump casing - Currently included as component type "pump" in LRA, Table 3.1.2-2, page 3.1-94.

• RV internal lower support plate - Currently included as part of "RVI Lower Core Support Structure (All RVI Stainless Steel Components)" in LRA, Table 3.1.2-2, page 3.1-78.

• RV internal lower support column - Currently included as part of "RVI Lower Core Support Structure (All RVI Stainless Steel Components)" in LRA Table 3.1.2-2, page 3.1-78.

'Baffle former plates - RVI baffle & former assembly - No time limited aging analysis (TLAAs) exist for this component, therefore no TLAA line is required.

Enclosure 1 PG&E Letter DCL-10-121 Page 29 of 29

2. With the exception of those listed below, the piping systems listed in Request for Additional Infor'mation 4.3-12: (1) are designed to ASME Class 2, 3, or ANSI B31.1 piping requirements, (2) are within the scope of license renewal, and (3) are subject to cumulative fatigue damage through the application of a stress range reduction factor. PG&E has evaluated the above list of piping systems in LRA Section 4.3.5.

Their inclusion in the AMR tables would only reference LRA Chapter 4.0 for the disposition through the inclusion of the phrase "Time Limited Aging Analysis evaluated for the period of extended operation" consistent with other generic aging lessons learned line items.

" Miscellaneous heating, ventilation, and air conditioning (HVAC) (Table 3.3.2-9) -

Not designed to ASME Class 2, 3, or ANSI B31.1 requirements

• Control Room HVAC (Table 3.3.2-10) - Not designed to ASME Class 2, 3, or ANSI B31.1 requirements

" Auxiliary Building HVAC (Table 3.3.2-11) - Not designed to ASME Class 2, 3, or ANSI B31.1 requirements

PG&E Letter DCL-10-121 Page 1 of 17 LRA Amendment 12 LRA Section RAI Table 4.3-2 4.3-4 Section 4.3.1.2 4.3-5 Table 3.1.2-1 4.3-5 Table 3.1.2-3 4.3-5 Table 3.1.2-4 4.3-5 Table 4.1-1 4.3-7 Section 4.3.2.1 4.3-7 Section 4.3.2.2 4.3-7 Section 4.3.2.3 4.3-7 Section 4.3.2.4 4.3-7 Section 4.3.4 4.3-7 Section A3.2.1.1 4.3-7 Section A3.2.1.2 4.3-7 Section A3.2.1.3 4.3-7 Section A3.2.1.4 4.3-7 Section A3.2.3 4.3-7 Section 4.3.1.2 4.3-5 Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 2 of 17 Table 4.3-2 DCPPUnits 1 and 2 Transient Cycle Count and 60-year Projections"" 0 Design Unit 1 Unit 2 Basis Limiting Transient Description Cycles, Analyzed Events Projected Events Projected FSAR Value ('1 (1984- Events (1985- Events Table 5.2-4 2008) for 60- for 60-2008) Years 2008) Years

14. Refueling(f NS 80 14 35 14 36

15. Charging and Letdown Flow Shutoff and Return NS 75 3/0 7 / 8(iv) 4/0 11 / 8 (g) to Service (Loop 4 I 3 )(0

16. Loss of Charging with Prompt Return NS 25 64/0 126/319) 54/0 118/3(g) to Service (Loop 4 N 3)(

17. Loss of Charging with Delayed NS 25 3/0 6/3(g) 3/0 6 3(g)

Returnto Service (Loop 4 / 3 )(')

18. Loss of Letdown with Prompt Return toSroice (oopr4 NS 250 60/0 124/25(! ) 45/0 94 / 2 5 (g) to Service (Loop 4/

3(f)

19. Loss of Letdown with Delayed NS 25 8/0 13 / 3(0) 10/0 22 / 3'g)

Return to Service (Loop 4 / 3 )(fý

20. Auxiliary Spray during Plant NS NS 78 146 54 102 Cooldown Upset Conditions 2-0-.21. Loss of Load (above 15%

Full Power), 100 19 5 18 10 Turbine Trip without Reactor Trip 24-22. LossofAll 50 3 1 2 1 3 Offsite Power 2,2L.23. Partial Loss of 100 3 1 3 3 8 Flow (1 RCP) 2-..24. ReactorTrip 500 88 58 100 48 83 from Full Power 24-.25.' Inadvertent RCS De-Pressurization NS 20 0 5 (h) 5 (Resulting in Reactor Trip)

Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 3 of 17 Table 4.3-2 DCPPUnits 1 and 2 Transient Cycle Count and 60-year Projections(i"ii)

Design Unit 1 Unit 2 Basis Limiting Transient Description Cycles, Analyzed Events Projected Events Projected FSAR Val Events (1985- Events Table 5.2-4 2008) for 60- 2008) for 60-2008)_Years Years 2-r26. Inadvertent Auxiliary Spray (differential 12 7 2 5 5 7 temperature

> 320°F) 26-27. Control Rod NS 80 0 5 1 2 Drop(0 2-7-.28. Inadvertent NS 60 0 5 0 5 ECCS ActuationN65 28.29. Excessive Feedwater Flow"' NS 30 0 1 0 1 2-9.30. Safety Injection into RCS Cold Leg NS 97 1 4 (') 0 4(v)

I High Head Safety Injection 30-.31. Inadvertent Accumulator NS 5 0 1 0 1 Blowdown(f_

3.4-.32. ,-1.2Dein20 Design 20 0 1 0 1 Earthquake (OBE)

Test Conditions 3-..33. Turbine Roll 10 10 5 8(vi) 6 9o)

Test 3-.34. Primary Side 10 5 1 20) 1 20)

Hydrostatic Test 34-.35. Secondary Side Hydrostatic Test 10 10 0 1 0 1 (each generator) 35-36. Primary Side 60 5 0 5 0 5 Leak Test 36-.37. Secondary Side 10 10 0 1 0 1 Leak Test 3-.38. Tube Leak 800 800 0 See Note k 0 See Note k Tests I / I Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 4 of 17 LRA Section 4.3.1.2 Present and Projected Status of Monitored Locations Projection Method Projected cycle counts were calculated using a dual linear projection of the historical results, except as noted in Table 4.3-2. For each event, two rates were determined; a long-term rate based on the entire history (i.e., the number of cycles since plant startup), and a short term rate (i.e., the incremental cycles over the last 10 years / 10 years). These two rates were combined using a weighted average:

Projection rate = [(LTW)*(Iong-term rate) + (STW)*(short-term rate)] / [(LTW) + (STW)].

The values of LTW (long-term weight) and STW (short-term weight) were determined on an event- OF ..... specific basis to reflect the most likely future behavior of that event or component. For most transients, the projection weighted the last 10 years more heavily based on the assumption that recent (short-term) history defines a trend which will likely continue into the future. For events that occurred infrequently, the projection increased the long-term weight since the recent history may have reflected isolated incidences rather than real trends.

These projections are intended to be a best estimate of the actual cycles expected.

They do not represent a revision of the design basis for the DCPP Units. The purpose is to demonstrate that the 50-year design numbers of transients are reasonable for 60 years. Future cycle count projections will be based on the actual accumulation history over the analysis period, adjusted on a component-specific basis by scaling factors to account for expected future operating conditions.

Section 3.1 PG&E Letter DCL-10-121 AGING MANAGEMENT OF REACTOR VESSEL, Page 5 of 17 INTERNALS, AND REACTOR COOLANT SYSTEM Table 3.1.2-1 Reactor Vessel, Internals,and Reactor Coolant System - Summary of Aging Management Evaluation -

Reactor Vessel and Internals Component Intended Aging Effect Type Fnton Material Environment Requiring Aging Management NUREG-1801 Table 1 Notes Type Function Management Program Vol. 2 Item Item Time Limited Aging RV Nozzle Carbon Plant Indoor Air Cumulative Analysis evaluated for IV.A2-20 3.1.1.01 C Support Pads SS Steel P Fatigue Damage the period of extended operation RVI Core Barrel DO, SH, Stainless Reactor Coolant Cumulative Time Limited Aging Analysis evaluated for IV.B2-31 3.1.1.05 A BArely SS Steel Fatigue Damage the period of extended Assembly operation Time Limited Aging RV Core SS Nickel Alloys Reactor Coolant Cumulative Analysis evaluated for IV.B2-31 3.1.1.05 Q Support Luqs Fatigue Damage the period of extended operation Section 3.1 PG&E Letter DCL-10-121 AGING MANAGEMENT OF REACTOR VESSEL, Page 6 of 17 INTERNALS, AND REACTOR COOLANT SYSTEM Table 3.1.2-3 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation -

Pressurizer Component Intended Aging Effect Aging Management NUREG-1801 Table 1 Function Material Environment Requiring Program Vol. tem Notes Type Management Program Vol. 2 Item Item Valve Support Time Limited Aging Bracket (Unit 2 SS Carbon Plant Indoor Air Cumulative Analysis evaluated for Il1.B1.1-12 3.5.1.42 A BctUtS Steel Fatigue Damage the period of extended only) operation Section 3.1 PG&E Letter DCL-10-121 AGING MANAGEMENT OF REACTOR VESSEL, Page 7 of 17 INTERNALS, AND REACTOR COOLANT SYSTEM Table 3.1.2-4 Reactor Vessel, Internals,and Reactor Coolant System - Summary of Aging Management Evaluation-Steam Generators Component Intended Aging Effect Aging Management NUREG-1801 Table I Material Environment Requiring Program Vo1. 2 1 Notes Type Function Management Program Vol. 2 Item Item SG Secondary Time Limited Aqingq Manway and PB Nickel Alloys Secondary Water Cumulative Analysis evaluated for IV.01-21 3.1.1.06 C Handhole (ext) Fatigue Damage the period of extended I Covers operation Time Limited Aging SG Feedwater DF Nickel Alloy Secondary Water Cumulative Analysis evaluated for IV.D1-21 3.1.1.06 C Bing (int) Fatigue Damage the period of extended operation Time Limited Aging SG Feedwater DF Carbon Secondary Water Cumulative Analysis evaluated for IV.D1-11 3.1.1.07 Q Ring Steel (ext) Fatigue Damage the period of extended _

operation Time Limited Aging Manway PB Stainless Reactor Coolant Cumulative Analysis evaluated for IV.D1-8 3.1.1.'10 C wPB Steel (ext) Fatigue Damage the period of extended Covers operation Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 8 of 17 Table 4.1-1 List of TLAAs TLAA Disposition Description Category(vii) Section Category

1. Reactor Vessel Neutron Embrittlement Analysis NA 4.2 Neutron Fluence Values 4.2.1 Pressurized Thermal Shock ii, iii 4.2.2 Charpy Upper-Shelf Energy ii 4.2.3 Pressure - Temperature Limits iii 4.2.4 Low Temperature Overpressure Protection iii 4.2.4

2. Metal Fatigue Analysis NA 4.3 DCPP Fatigue Management Program NA 4.3.1 ASME Section III Class A Fatigue Analysis of Vessels, NA 4.3.2 Piping, and Components Reactor Pressure Vessel, Nozzles, and Studs ii, iii 4.3.2.1 Reactor Vessel Closure Head and Associated 4.3.2.2 Components Reactor Coolant Pump Pressure Boundary i, iiii 4.3.2.3 Components Pressurizer and Pressurizer Nozzles i, ii, iii 4.3.2.4 Steam Generator ASME Section III Class 1, Class 2 Secondary Side, and Feedwater Nozzle Fatigue i 4.3.2.5 Analyses and Fatigue Qualification Tests Absence of TLAA for Reactor Coolant System NA 4.3.2.6 Boundary Valves Reactor Coolant Pressure Boundary Piping NA 4.3.2.7 Absence of Supplemental Fatigue Analysis TLAAs in Response to Bulletin 88-08 for Intermittent Thermal Cycles due to Thermal-Cycle-Driven Interface Valve Leaks and Similar Cyclic Phenomena Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and iii 4.3.2.9 Stratification Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 9 of 17 Table 4.1-1 List of TLAAs TLAA Category Description Disostigony)Seto Category~vi eto Disposition Absence of a TLAA for Thermal Embrittlement of Cast Austenitic Stainless Steel (CASS) Reactor Coolant NA 4.3.2.10 Pumps, Absence of a Cumulative Fatigue Usage Factor TLAA to Determine High Energy Line Break (HELB) NA 4.3.2.11 Locations TLAAs in Fatigue Crack Growth Assessments and Fracture Mechanics Stability Analyses for Leak-Before- 4.3.2.12 Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures Fatigue Analyses of the Reactor Pressure Vessel Internals iii 4.3.3 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety ii, iii 4.3.4 Issue 190)

Assumed Thermal Cycle Count for Allowable Secondary i 4.3.5

-Stress Range Reduction Factor in ANSI. B3!.1 Piping Fatigue Design and Analysis of Class IE Electrical 4.3.6 Raceway Support Angle Fittings for Seismic Events Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 10 of 17 LRA Section 4.3.2.1, page 4.3-15 Disposition: ValidatielRevision, 10 CFR 54.21(c)(1)(ii) and Aging Management, 10 CFR 54.21(c)(1)(iii)

ValldatioRevision As shown in Table 4.3-3, the usage factors for all RPV components, with the exception of the RPV studs and core support pads, calculated in this analysis remain significantly below 1.0 (i.e., do not exceed 0.6, when projected to 60 years). All RPV components, with the exception of the RPV studs and core support pads, will be valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1 )(ii).

Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 11 of 17 LRA Section 4.3.2.2, pages 4.3-16 and 4.3-17 Disposition: ValidationRevision, 10 CFR 54.21(c)(1)(ii)

Validation - RRVCH The Unit 1 and 2 replacement reactor vessel heads including the RRVCHs, CRDMs, CETNAs, and thermocouple nozzles will be analyzed for a 50-year design life, which will extend beyond the period of extended operation. Therefore the fatigue analyses the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).

Vafdatien Revision - Thermocouple Column with Low Design Basis Usage Factors The current fatigue analyses of the thermocouple column demonstrate that the maximum 40-year usage factor is 0.29. If multiplied by 1.5 (60/40) to account for the 60-year period of extended operation, these results do not exceed 0.6, providing a large margin to the code acceptance criterion of 1.0. The analyses of these components are therefore valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(ii).

Enclosure 2 Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS

.Page 12 of 17 LRA Section 4.3.2.3, page 4.3-20 Disposition: Validation, 10 CFR 54.21(c)(1)(i); Revision, 10 CFR 54.21(c)(1)(ii); and Aging Management, 10 CFR 54.21 (c)(1)(iii).

Validation - Hydraulic Nuts and Studs The Unit 1 RCP 1-2 hydraulic nuts and studs were installed in 2005 with a 50-year design life, which will extend beyond the period of extended operation. Therefore, the fatigue analyses for the hydraulic nuts and studs will remain valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1 )(i).

IVaidati94 Revision - Thermal BarrierFlange and Main Flange Thermowell The design basis fatigue usage in the thermal barrier flange is a negligible, 0.0002. The thermal barrier flange design CUF was multiplied by 1.5 (60/40) resulting in a CUF of 0.0003 for 60 years of operation.

The design basis analysis qualified the main flange thermowell for greater than 106 cycles, indicating an alternating stress intensity that is less than the endurance limit.

The increase in design life from 40 years to 60 years does not affect this basis for the safety determination.

Therefore, the fatigue analyses of the thermal barrier flange and main flange thermowell will remain valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(ii).

Enclosure 2 Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 13 of 17 LRA Section 4.3.2.4, page 4.3-24 Disposition: Validation, 10 CFR 54.21 (c)(1)(i); Revision, 10 CFR 54.21 (c)(1)(ii);

and Aging Management, 10 CFR 54.21(c)(1)(iii)

Validation - Unit 2 Relief Valve Support Bracket, Including Permitted Relief Valve Operating Cycles The analysis of the Unit 2 relief valve support bracket determined the partial usage factor due to loads required by the design specification is much less than 0.1.

Maintaining the usage factor below 1.0 is controlled by the permitted number of valve operating cycles. However, the limit is above 9,000 operations, far in excess of any expected, in any foreseeable design life. The fatigue analysis of the Unit 2 relief valve support bracket is therefore valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).

I VafidatenRevision - PressurizerSubcomponents with Projected 60-Year Usage FactorsLess Than 0.6 As shown in Table 4.3-6, the projected 60-year fatigue usage factors of some subcomponents remain significantly below 1.0 (i.e. do not exceed 0.6, when projected to 60 years). The analyses of these subcomponents are therefore valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1 )(ii).

Enclosure 2 Section 4 PG&E Letter DCL-10-121 TIME-LIMITED AGING ANALYSIS Page 14 of 17 LRA Section 4.3.4, page 4.3-45 Disposition: ValidationlRevision, 10 CFR 54.21(c)(1)(ii); and Aging Management, 10 CFR 54.21 (c)(1)(iii)

IVadatioRRevision, 10 CFR 54.21(c)(1)(ii)

As shown in Table 4.3-8, the evaluation of fatigue effects in three of the NUREG/CR-6260 locations, reactor vessel shell to lower head junction, reactor vessel inlet nozzles, and RHR line tee, has demonstrated that the EAF CUF values will remain sufficiently below 1.0, i.e., less than 0.5. If multiplied by 60/50 to account for the period of extended operation, these results do not exceed 0.6, providing a large margin to the code acceptance criterion of 1.0. The evaluation of fatigue effects in these locations has been validated and projected to the end of the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(ii).

Appendix A PG&E Letter DCL-10-121 Final Safety Analysis Report Supplement Page 15 of 18 A3.2.1.1 Reactor Pressure Vessel, Nozzles, and Studs The DCPP Unit 1 reactor pressure vessel is designed to ASME Code, Section II1., 1965 Edition through the Winter 1966 Addenda. The DCPP Unit 2 reactor pressure vessel is designed to ASME Section III 1968 Edition.

Pressure-retaining and support components of the reactor pressure vessel are subject to an, ASME Boiler and Pressure Vessel Code Section III fatigue analysis. This original fatigue analysis has been updated to incorporate redefinitions of loads and design basis events, operating changes, replacement steam generators, and minor modifications using the 50-year design basis number of transients.

The usage factors for all reactor pressure vessel components, with the exception of the RPV studs and core support pads, remain below 1.0 when projected to 60. years. All RPV components, with the exception of the RPV studs and core support pads, will be valid .for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(ii).

The Metal Fatigue of Reactor Coolant Pressure Boundary program described in Section A2.1 will ensure that the fatigue analyses for RPV studs and core support pads remain valid, or that appropriate reevaluation or other corrective measures maintain the design and licensing basis. Action limits will permit completion of corrective actions before the design basis number of events is, exceeded, and before the cumulative usage factor exceeds the code limit of 1.0. Therefore, effects of fatigue in the reactor pressure vessel pressure boundary and its supports will be managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

A3.2.1.2 Reactor Vessel Closure Heads and-Associated Components The reactor pressure boundary components associated with the reactor vessel closure head are the control rod drive mechanisms (CRDM) pressure housings, core exit thermocouple nozzle assemblies (CETNAs), thermocouple nozzles, and thermocouple columns. The Units 1 and 2 CRDMs pressure housings, the CETNAs, and the thermocouple nozzles will be replaced with the replacement reactor vessel closure heads (RRVCHs). The RRVCHs, CRDM pressure housings, CETNA, and thermocouple nozzles will be designed to ASME Code,Section III. The Unit 1 and 2 RRVCHs, CRDMs, CETNAs, and thermocouple nozzles will be analyzed for a 50- year design life, and therefore will remain valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(i).

The only reactor pressure boundary components associated with the reactor vessel closure head that will not be replaced are the thermocouple columns. These components were designed to the ASME Code, Section IIl. The current fatigue analyses of the thermocouple column demonstrate a large margin to the code acceptance criterion of 1.0. The analyses of these components are therefore valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ij).

Appendix A PG&E Letter DCL-10-121 Final Safety Analysis Report Supplement Page 16 of 18 A3.2.1.3 Reactor Coolant Pump Pressure Boundary Components There are four Model 93A Reactor Coolant Pumps (RCPs) for each reactor (one pump per coolant loop). The RCP design reports demonstrate that the pressure components satisfy all the Class A requirements of the ASME Code,Section III, 1968 Edition through the Winter 1970 Addenda.

Thermal BarrierFlange and Main Flange Thermowell Fatigue in the thermal barrier flange and main flange thermowell was shown to be negligible based on the low CUF and the high number of allowable cycles. Increasing the 40-year design life results from the generic stress reports by a factor of 1.5 to account for a 60-year design life would not change this determination. Therefore the fatigue analysis of the thermal barrier flange and main flange thermowell will remain valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).

Hydraulic Nuts and Studs Hydraulic nuts and studs were installed on Unit 1 RCP 1-2 in 2005. These components were analyzed with a 50-year design life, which will extend beyond the period of extended operation. Therefore the fatigue analyses for the hydraulic nuts and studs will remain valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).

The Metal Fatigue of Reactor Coolant Pressure Boundary program described in Section A2.1 ensures that either the assumed numbers of design cycles or transient events used by the RCP design documents for the Locating Slot, Main Flange Bolts, and Seal Housing Penetrations and Bolts are not exceeded, or that appropriate reevaluation or other corrective action is taken if a design basis number of events is approached.

Action limits will permit completion of corrective actions before the design basis number of events is exceeded, and before the cumulative usage factor exceeds the code limit of 1.0. Effects of fatigue in the reactor coolant pump pressure boundaries will be managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

A3.2.1.4 Pressurizer and Pressurizer Nozzles The pressurizers and their integral support skirts are designed to ASME Section III, 1965 Edition, with Addenda through Summer of 1966, as ASME Section III Class A components.

Appendix A PG&E Letter DCL-10-121 Final Safety Analysis Report Supplement Page 17 of 18 Unit 2 Relief Valve Support Bracket The analysis of the Unit 2 relief valve support bracket determined the partial usage factors due to loads required by the design specification and due to relief valve operation. Assuming an increase in usage factor due to design specification loads, proportional to the increase in licensed operating period, limits the fatigue usage available for relief valve operation. However the limiting number of valve operating cycles is far in excess of any expected in any foreseeable design life. The fatigue analysis of the Unit 2 relief valve support bracket is therefore valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1 )(i).

PressurizerSubcomponents with Projected 60-Year Usage FactorsLess Than 0.6 The projected 60-year fatigue usage factors of some pressurizer subcomponents remain significantly below 1.0 (i.e., do not exceed 0.6 when projected to 60 years). The analyses of these subcomponents are therefore valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).

Appendix A PG&E Letter DCL-10-121 Final Safety Analysis Report Supplement Page 18 of 18 A3.2.3 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)

DCPP addressed Generic Safety Issue - 190 review requirements by assessing the environmental effect on fatigue at the NUREG/CR-6260 sample locations for older-vintage Westinghouse plants. NUREG/CR-6260 identifies seven sample locations for older-vintage Westinghouse plants:

" Reactor vessel shell and lower head

" Reactor vessel inlet nozzles

" Reactor vessel outlet nozzles

" Pressurizer surge line (hot leg nozzle safe end)

" Charging system nozzle

" Safety injection system nozzle

" Residual heat removal system piping.

The evaluation of fatigue effects in three of the NUREG/CR-6260 locations, reactor vessel shell to lower head junction, reactor vessel inlet nozzles, and RHR line tee, has demonstrated that the CUF values will remain sufficiently below 1.0 using the maximum applicable Fen values to validate them for the period of extended operation. If multiplied by 60/50 to account for the 60-year period of operation, these results do not exceed 0.6, providing a large margin to the code acceptance criterion of 1.0. The evaluation of fatigue effects in these locations has been validated and projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1 )(ii).