ML17138B115

10 CFR 50.55a Request for Approval of Alternative - RI-ISI-2

ML17138B115
Person / Time
Site:	Diablo Canyon
Issue date:	05/18/2017
From:	Pacific Gas & Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17139C628	List: DCL-17-048, Diablo Canyon, Units 1 and 2 - Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section Xi Examination Requirements for Class 1 and 2 Piping Welds, Request for R DCL-17-048, Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Examination Requirements for Class 1 and 2 Piping Welds, Request for RI-ISI-1 ML17138B115
References
DCL-17-048
Download: ML17138B115 (49)

Text

Enclosure 2 PG&E Letter DCL-17 -048 Diablo Canyon Power Plant Unit 2 10 CFR 50.55a Request for Approval of Alternative - RI-ISI-2

Enclosure 2 PG&E Letter DCL-17 -048 Diablo Canyon Power Plant Unit 2 10 CFR 50.55a Request Number RI-ISI-2 Proposed Alternative In Accordance with 10 CFR 50.55a(z){1)

-Alternative Provides Acceptable Level of Quality and Safety-I. ASME Code Components Affected Code Class 1 and 2 piping welds previously subject to the requirements of American Society of Mechanical Engineers (ASME)Section XI, Table IWB-2500-1, Examination Categories B-F* and B-J, and Table IWC-2500-1, Examination Categories C-F-1 and C-F-2, are affected.

II. Applicable Code Edition and Addenda The Diablo Canyon Power Plant (DCPP) Unit 2 lnservice Inspection (lSI) program for the fourth lSI interval is based on the 2007 Edition of ASME Section XI through the 2008 Addenda.

Ill. Applicable Code Requirement The selection of Code Class 1 and Code Class 2 pipe welds to be examined in the fourth inspection interval is required to be prescriptively determined in accordance with Table IWB-2500-1, Examination Categories B-F* and B-J, and Table IWC-2500-1, Examination Categories C-F-1 and C-F-2.

IV. Reason For Request The continued use of a risk-informed process as an alternative for the selection of Class 1 and Class 2 piping welds for examination is requested for the fourth lSI Interval of Unit 2. Use of the risk-informed selection process has been shown to reduce the core damage frequency and large early release frequency when compared to the prescriptive deterministic selection method.

• Note that although Examination Category 8-F welds are included in the RI-ISI program for other damage mechanisms, Alloy 600/82/182 examinations in the Third Interval were conducted per Code Cases N-722-1 and N-770-1. In the fourth interval, these examinations will be performed in accordance with the versions of the applicable Code Cases that are referenced in the published version of 10 CFR 50.55a.

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Enclosure 2 PG&E Letter DCL-17 -048 V. Proposed Alternative and Basis for Use As an alternative to the Code Requirement, a risk-Informed process will continue to be used for selection of Class 1 and Class 2 piping welds for examination.

The DCPP Unit 2 lSI program for examination of Class 1 and Class 2 piping welds is currently in accordance with a risk-informed process developed and based on EPRI TR-112657, Revision B-A, with identified differences and with additional guidance taken from ASME Code Case N-578. In 2001, DCPP submitted a request for alternative in PG&E letter DCL-01-015, "Relief Request for Application of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Examination Requirements for Class 1 and 2 Piping Welds," dated Feb'ruary 16, 2001 (Examination Categories B-F, B-J, C-F-1, and C-F-2) inservice inspections to implement a risk-informed inservice inspection (RI-ISI) program. The NRC published a safety evaluation authorizing the use of the RI-ISI program for the second 10-year lSI interval for DCPP Units 1 and 2. Both the original RI-ISI submittal and the resultant NRC Safety Evaluation call for a periodic review and update of the program. An update was performed for the end of the third period of the second interval. Based on that update, another request for alternative for the third lSI interval was submitted in PG&E Letter DCL-12-007, "Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Examination Requirements for Class 1 and 2 Piping Welds," dated January 20, 2012. DCL-12-007 was supplemented by PG&E Letter DCL-12-084,

• "Response to NRC Request for Additional Information Regarding Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Examination Requirements for Class 1 and 2 Piping Welds," dated September 6, 2012. This request was approved for the entire third interval. The resultant program was implemented for the third interval, and was reviewed and updated after the first, second and third periods of the third interval.

In accordance with NEI 04-05 (April 2004), the following aspects were considered during the reviews:

• Plant Examination Results

• Piping Failures

-Plant Specific Failures

-Industry Failures

• PRA Updates

• Plant Design Changes

-Physical Changes

-Programmatic Changes

-Procedural Changes

• Changes in Postulated Conditions

-Physical Conditions

-Programmatic Conditions 2

Enclosure 2 PG&E Letter DCL-17 -048 The updated program resulting from these reviews is the subject of this proposed alternative.

In accordance with the guidance provided by NEI 04-05, a table is provided as identifying the number of welds added to and deleted from the previously approved RI-ISI program . The changes from the previous program are attributable to the specific issues(s) identified in each review:

During the review after the first period of the third lSI interval, the following issues were identified:

1. In the chemical and volume control system (CVCS), valves CVCS-2-8372A, B, C, and CVCS-2-8367 A, B, C, and CVCS-2-8479-A, B were replaced. In the reactor coolant system (RCS), pressurizer nozzle safe end welds received weld overlays. Multiple welds were deleted, added, or renamed as a result of steam generator (SG) replacement, centrifugal charging pump replacement, and positive displacement pump replacement. As a result, there were multiple changes to the weld population.

2. Based on a change to ASME Section XI Code criteria, the 4-inch nominal pipe size (NPS) Class 2 auxiliary feedwater (AFW) lines from the level control valves to their respective connections to the four main feedwater lines were added to the RI-ISI Program.

3. Six weld overlays were installed as a result of implementation of the RCS Alloy 600 Program. Due to the proximity of adjacent welds, these overlays actually overlaid 12 welds.

During the review after the second period of the third lSI interval, the following issues were identified:

1. The DCPP probabilistic risk assessment (PRA) model used to evaluate the consequences of pipe rupture for the previous RI-ISI update was Model DC01 dated June 2006. Model DC01 was still the model of record during the period under evaluation . As such, there was no change required to any consequence analysis or to the upper bound conditional core damage probability (CCDP) or large early release probability (LERP). However, the model of record (MOR) changed to DC02 in November of 2012. PG&E decided to proactively reflect this change as part of the Interval 3, Period 2 evaluation. For this model the core damage frequency (CDF) is 6.91 E-05/yr and large early release frequency (LERF) is 3.17E-06/yr. Maximum CCDP used as the upper bound in the risk impact analysis is 3.98E-02 associated with Consequence Cases CVCS-1, 3

Enclosure 2 PG&E Letter DCL-17 -048 RCS-1, and Sl-3. The update to the PRA model resulted in the following changes in consequence rankings:

ConsequenceiD DC01 DC02 Change in Rank Rank Consequence Rank ACC02A M H Medium to High ACC02B M H Medium to High ACC02C M H Medium to High ACC02D M H Medium to High CS01 M H Medium to High CS02 M H Medium to High CS03A M H Medium to High CS04A M H Medium to High CS03B M H Medium to High CS04B M H Medium to High CVCS05B M L Medium to Low CVCS07 M L Medium to Low CVCS08 M L Medium to Low CVCS09 M L Medium to Low

2. During the first period of the third lSI interval, the RI-ISI Program was subjected to an extensive review and verification. During the second period of the third lSI interval, the updated risk ranking, summary, and matrix were used to reflect the resulting findings and reconciliations.

3. During the element selection process, it was noted that the four welds in CVCS Risk Category 5a and subject to thermal stratification, cycling, and striping (TASCS) were all single-sided welds and none could be properly examined.

Since only one weld was required to be inspected, a weld in the same system with the same degradation mechanism, but a higher Risk Category, was selected as a substitute. In Unit 2, S6-50-3-WIB-186 was selected.

During the review of the third period of the third lSI interval, the following issues were identified:

1. During the Unit 2 Nineteenth Refueling Outage, an indication was found in Class 1 Weld WIB-245. Several crack growth and degradation mechanisms were investigated as part of that evaluation since the cracking did not conform to any known industry operating experience. The specific mechanisms considered include thermal shock from cyclic swirl penetration and cyclic thermal stratification of the unisolable horizontal pipe section and stress corrosion cracking. Temperature monitoring was conducted to further refine the analysis.

Vibration was ruled out as a possible cause based on inspections performed by 4

Enclosure 2 PG&E Letter DCL-17 -048 PG&E, which concluded that physical evidence indicative of excessive vibration was not present.

Because the evaluation could neither rule out nor specify the specific flaw growth mechanism, a conservative analysis was performed combining the effects of two different flaw growth mechanisms comprised of fatigue crack growth (FCG) and stress corrosion cracking (SCC), for justification of continued operation for an additional cycle. PG&E is not attributing sec with respect to risk-informed lSI weld inspections due to uncertainty regarding the flaw growth mechanism.

2. The DCPP PRA was updated to Model DC03 in July 2015. In Model DC03, the total CDF is 5.52E-05/yr and the total LERF is 5.73E-06/yr. The maximum CCDP used as upper bound in the risk Impact analysis is 1.74E-02 and the maximum conditional large early release probability (CLERP) is 7.03E-03, both associated with Consequence Cases CVCS-1, RCS-1, and Sl-3. The update in PRA model resulted in the following changes in consequence rankings:

Change in ConsequenceiD DC02 Rank DC03 Rank Consequence Rank ACC02A H M High to Medium ACC02B H M High to Medium ACC02C H M High to Medium ACC02D H M High to Medium CS01 H M High to Medium CS02 H M High to Medium CS03A H M High to Medium CS04A H M High to Medium CS03B H M High to Medium CS04B H M High to Medium CVCS01B M L Medium to Low CVCS02B M L Medium to Low RHR01 L M Low to Medium RWST02A-PEN M H Medium to High RWST02B-PEN M H Medium to High RWST03A M H Medium to High RWST03B M H Medium to High SI01 M H Medium to High SI02 M H Medium to High SI03A M H Medium to High 81038 M H Medium to High All issues identified in the periodic reviews have been incorporated into the risk ranking, summary, and matrix. Limits are imposed by the EPRI methodology to ensure that the change in risk of implementing the RI-ISI program meets the requirements of 5

Enclosure 2 PG&E Letter DCL-17-048 Regulatory Guides 1.174 and 1.178. The EPRI criterion requires that the cumulative change in CDF and LERF be less than 1E-07 and 1E-08 per year per system, respectively. A new risk impact analysis was performed, and the revised program continues to represent a risk reduction when compared to the last deterministic Section XI inspection program . The revised program represents an overall reduction of plant risk of 4.96E-08 inCDF and 2.00E-08 in LERF.

As indicated in the following table, this evaluation has demonstrated that unacceptable risk impacts will not occur for any system from implementation of the RI-ISI program regardless of whether the enhanced probability of detection (POD) is credited for the RI-ISI examinations.

Unit 2 Risk Impact Results ARiskcoF ARiskLERF System w/POD w/o POD w/POD w/o POD ReS -1.66E-08 1.48E-09 -6.71 E-09 5.98E-10 eves -7.13E-09 -4.35E-09 -2.88E-09 -1.76E-09 SIS -1.59E-08 -8.94E-09 -6.43E-09 -3.62E-09 RHRS -9.65E-09 -4.78E-09 -3.90E-09 -1.93E-09 ess 2.51 E-12 2.51E-12 2.51E-13 2.51E-13 RWST -2 .61 E-10 -2.61 E-10 -1.05E-10 -1 .05E-10 eew O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO FWS 3 .80E-13 6 .20E-13 3.80E-14 6 .20E-14 MSS 7.50E-14 7.50E-14 7.50E-15 7.50E-15 AFW 1.65E-13 2.45E-13 1.65E-14 2.45E-14 Total -4.96E-08 -1.68E-08 -2.00E-08 -6.82E-09 The following augmented inspection programs were considered during the RI-ISI application:

• The augmented examination program for flow accelerated corrosion (FAC) per NRC Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning," dated May 2, 1989, is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RI-ISI program.

• The augmented examinations for thermal fatigue in non-isolable reactor coolant system branch lines are performed in accordance with EPRI Materials Reliability Program document, MRP-146, which is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RI-ISI program.

• The augmented visual examinations for pressure retaining welds in Class 1 components fabricated with Alloy 600/82/182 materials are performed in accordance with Code Case N-722-1, which is relied upon to manage the damage mechanism of primary water stress corrosion cracking (PWSCC) but is not otherwise affected or changed by the RI-ISI program.

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Enclosure 2 PG&E Letter DCL-17-048

• The augmented examinations and acceptance standards for Class 1 piping and vessel nozzle butt welds fabricated with UNS N06082 or UNS W86182 weld filler metal are performed in accordance with Code Case N-770-1 which is relied upon to manage the damage mechanism of PWSCC but is not otherwise affected or changed by the RI-ISI program. Note that welds selected for examination in accordance with Code Case N-770-1 are considered as part of the RI-ISI population such that they are evaluated for other potential degradation mechanisms. However, they are excluded from selection under the RI-ISI Program. In the fourth interval these examinations will be performed in accordance with the version of Code Case N-770 that is referenced in the published-version of 10 CFR 50.55a. This is expected to be Code Case N-770-2 per the Notice of Proposed Rulemaking dated September 18, 2015.

The RI-ISI program is a living program requiring feedback of new relevant information to ensure the appropriate identification of high safety significant piping locations. As a minimum, risk ranking of piping segments will be reviewed and adjusted on an ASME period basis. In addition, significant changes may require more frequent adjustment as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant specific feedback.

The risk-informed process continues to provide an adequate level of quality and safety for selection of the Class 1 and Class 2 piping welds for examination. Therefore, pursuant to 10 CFR 50.55a(z)(1 ), PG&E requests that the proposed alternative be authorized.

VI. PRA Quality The PRA Quality Assessment is provided in Attachment 2.

VII. Duration of Proposed Alternative The alternative will be used for DCPP Unit 2 until the end of that unit's fourth 10-year lSI Program inspection interval, subject to the review and update guidance of NEI 04-05.

The fourth inspection interval is currently scheduled to end on March 13, 2026.

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Enclosure 2 Attachment 1 PG&E Letter DCL-17 -048 DCPP Unit 2 - Inspection Location Selection Comparison Between Previously Approved and Revised Rl-151 Program by Risk Category Previously Approved Updated Risk Failure Potential 1 Consequence Code {Third Interval) {Fourth Interval)

System( )

Rank Category Weld Other( 2 ) Weld Other( 2 )

Category Rank OMs Rank RI-ISI RI-ISI Count Count TASeS, 9(3) 6(3) 9(3) 5(3)

ReS 2 High High Medium B-J TT ReS 2 High High TASeS Medium B-J 10 4 10 4 High TT Medium 1(4) 0(4) 1(4) 0(4)

ReS 2 (2) High B-F (High) / (PWSee) (Medium)

ReS 2 High High TT Medium B-J 13 0 13 0 TASeS, eves 2 High High Medium B-J 5 3 5 2 TT eves 2 High High TT Medium B-J 3 1 3 0 SIS 2 High High TT Medium B-J 18 6 18 5 RHR 2 High High TASeS Medium e-F-1 11 3 11 3 Medium None Low 0(4) 0(4) 13(4) 0(4)

ReS 4 (2) High B-F (High) (PWSee) (Medium)

B-F 21 2 8 0 ReS 4 Medium High None Low B-J 277 34 285 31 eves 4 Medium High None Low B-J 92 11 92 10 eves 4 Medium High None Low e-F-1 21 2 22 0 SIS 4 Medium High None Low B-J 30 4 30 11 SIS 4 Medium High None Low e-F-1 68 7 131 6 RHR 4 Medium High None Low e-F-1 175 18 175 18 RWST 4 Medium High None Low e-F-1 45 5 117 12 eew 4 Medium High None Low e-F-2 12 2 12 2 Al-1

Enclosure 2 Attachment 1 PG&E Letter DCL-17 -048 DCPP Unit 2 - Inspection Location Selection Comparison Between Previously Approved and Revised Rl-151 Program by Risk Category Previously Approved Updated Risk Failure Potential Consequence Code {Third Interval) {Fourth Interval)

System( 1l

- Rank Category Weld Weld Category Rank OMs Rank RI-ISI Other( 2l RI-ISI Other( 2 )

Count Count TASeS, eves 5a Medium Medium Medium B-J 2 1 0 0 TT eves 5a Medium Medium TT Medium B-J 2 0 0 0 SIS 5a Medium Medium IGSee Medium B-J 13 2 13 2 SIS 5a Medium Medium TASCS Medium e-F-1 4 0 4 1 ReS 6a Low Medium None Low B-J 3 0 3 0 eves 6a Low Medium None Low B-J 8 0 8 0 I eves 6a Low Medium None Low e-F-1 673 0 0 0 SIS 6a Low Medium None Low B-J 134 0 134 0 SIS 6a Low Medium None Low e-F-1 160 0 68 0 RHR 6a Low Medium None Low B-J 20 0 20 0 RHR 6a Low Medium None Low e-F-1 85 0 85 0 ess 6a Low Medium None Low e-F-1 72 0 72 0 RWST 6a Low Medium None Low e-F-1 72 0 0 0 TASeS, eves 6b Low Low Medium B-J 0 0 2 0 TT eves 6b Low Low TT Medium B-J 52 0 54 0 SIS 6b Low Low IGSee Medium B-J 7 0 7 0 AFW 6b Low Low TT Medium e-F-2 15 0 15 0 Low TASeS Medium FWS 6b (5b) Low e-F-2 28 0 32 0 (Medium) (FA e) (High)

Al-2

Enclosure 2 Attachment 1 PG&E Letter DCL-17 -048 DCPP Unit 2 - Inspection Location Selection Comparison Between Previously Approved and Revised Rl-151 Program by Risk Category I Previously Approved Updated Risk Failure Potential I 1 Consequence Code (Third Interval) (Fourth Interval)

System( )

Rank Category Weld Weld Category Rank OMs Rank RI-ISI Other( 2 l RI-ISI Other( 2 l Count Count RCS 7a Low Low None Low 8-J 13 0 13 0 eves 7a Low Low None Low C-F-1 0 0 748 0 I I

SIS 7a Low Low None Low 8-J 216 0 216 0 SIS 7a Low Low None Low C-F-1 9 0 34 0 css 7a Low Low None Low C-F-1 12 0 12 0 I MSS 7a Low Low None Low C-F-2 118 0 122 0 AFW 7a Low Low None Low c~F-2 130 0 130 0 Low None Low FWS 7a (5b)

(Medium)

Low (FAG) (High)

C-F-2 37 0 37 0 Notes

1. Systems were described in Table 3.1-2 of the original submittal (PG&E Letter DCL-01-015, dated February 16, 2001 ), with the exception of AFW. This ASME Code Class 2 system consists of 145 elements.

2. The column labeled "Other" is generally used to identify augmented inspection program locations that are credited beyond those locations selected per the RI-ISI process, as addressed in Section 3.6.5 of EPRI TR-112657, Rev. 8-A. This option was not applicable for the DCPP RI-ISI application. The "Other" column has been retained in this table solely for uniformity purposes with other RI-ISI application template submittals.

3. One of the elements selected for RI-ISI is the surge line elbow and is not counted as part of the weld count.

4. The examinations for these welds are performed in accordance with Code Case N-770-1, which is relied upon to manage the damage mechanism of PWSCC but is not otherwise affected or changed by the RI-ISI program . Note that welds selected for examination in accordance with Code Case N-770-1 are considered as part of the RI-ISI population such that they are evaluated for other potential degradation mechanisms. However, they are excluded from selection under the RI-ISI program. In the fourth interval, these examinations will be performed in accordance with the version of Code Case N-770 that is referenced in the published version of 10 CFR 50.55a. For the fourth interval, these welds have been re-categorized in the RI-ISI application for ease of identification.

Al-3

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Attachment 2 PRA Technical Adequacy for RI-ISI Application As discussed in the NRC safety evaluation of EPRI TR 1021467 and PG&E's response to RAI Question #7 for approval of the RI-ISI third interval as documented in PG&E Letter DCL-12-84, the impact of the external event PRAs do not significantly impact the RI-ISI application. Therefore the following DCPP PRA development history and technical adequacy is focused on the Internal Events and Internal Flooding PRAs.

A.1 History of DCPP PRA Model Development The current DCPP PRA model is based on the original 1988 Diablo Canyon PRA (DCPRA -1988) model, developed as part of the Long-Term Seismic Program (LTSP). The DCPRA -1988 was a full-scope Level1 PRA that evaluated internal and external events. The NRC reviewed the LTSP and issued Supplement No. 34 to NUREG-0675 in June 1991, accepting the DCPRA-1988. Brookhaven National Laboratory performed the primary review of the DCPRA-1988 for the NRC; their review is documented in NUREG/CR-5726.

The DCPRA-1988 was subsequently updated to support the Individual Plant Examination in 1991 and the Individual Plant Examination for External Events in 1993. Since 1993, several other updates have been made to incorporate plant and procedure changes, update plant-specific reliability and unavailability data, and to improve the fidelity of the model.

At the time the fourth RI-ISI consequence case ranking evaluation process started, the MOR was DC03. DC03 incorporated the resolution of 2012 Internal Events and Internal Flooding Peer Review facts and observations (F&Os) along with a routine data update. The fourth RI-ISI consequence case ranking is based on quantitative risk insights from MOR DC03. The latest MOR is DC03A which was updated in 2016 and incorporates Westinghouse safe shutdown reactor coolant pump (RCP) seal modeling into the internal events model. The DC03A update is not expected to impact the results of the consequence case ranking because the changes made in DC03A did not significantly influence the CCDPs, and initiating event frequencies used in the RI-ISI evaluation.

A.2 Internal Events and Internal Flooding PRA Peer Review The DCPP Internal Events and Internal Flooding PRA had a full scope peer review in accordance with NEI guidance. This review was conducted in December, 2012.

The peer review was done in accordance with Capability Category II requirements of A2-1

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 the ASME/ANS RA-Sa-2009 Standard as endorsed by RG 1.200, Revision 2, with the full consideration of NRC regulatory positions described in Appendix A, B, and C.

The peer review found the Internal Events PRA and Internal Flooding model to be technically adequate. The results of this peer review, including (F&O) resolutions and impact on this RI-ISI alternative request submittal, are summarized in Table A-1 for Interval Events and Table A-2 for Flooding.

A.3 Review of Modeling Uncertainties Table A4-2 in PRA Calculation C.1 0 Revision 7, "PRA Technical Adequacy," dated March 2016, provides a list potential modeling uncertainties and their characterization. The review of this table identified no key modeling uncertainty that could impact either the consequence analysis or risk ranking requiring changes to the model or sensitivity analysis.

A.4 PRA Maintenance and Upgrade The PG&E risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the DCPP risk management program and associated procedures.

These procedures delineate the responsibilities and guidelines for maintaining the PRA models at DCPP.

A.5 Conclusion DCPP Internal Events and Internal Flooding PRAs have been developed, refined, and maintained to reflect the as-built/as-operated condition of the plant per applicable industry guidance documents and PG&E administrative procedures. The Internal Events and Flooding PRAs have been peer reviewed to the latest PRA standard as endorsed by RG 1.200, Revision 2. All F&Os from the peer reviews were satisfactorily resolved and there is no open issue that could impact the results of this analysis.

DCPP Internal Events and Internal Flooding PRAs are technically adequate to support the RI-ISI alternative request.

A2-2

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level IE-A5 IE-A5-01 (Systematic F There is no evidence in the This F&O has been resolved by additional reviews; no review of each system) Closed documentation of a systematic new or changed initiating events were identified . Each evaluation of every system to system was screened for potential initiating events. If a IE-A5 not met assess the possibility of an system did not screen, it was then reviewed to confirm initiating event occurring due to that a bounding or representative initiating event is failure of the system. already modeled in the PRA. An interview with an Operations representative was conducted to confirm the system screening and to discuss low power or non-power operations for each system.

This supporting requirement (SR) is judged to now be met at capability category II, based on the use of a structured approach for evaluation of each system for initiating event potential.

IE-A7 IE-A7-01 (Events which F Closed The identification of initiating This F&O has been resolved by additional reviews; no occurred other than at- events does not include new or changed initiating events were identified . Are-power) consideration of events occurring review of plant information in the Twice-Daily Shift during low-power or shutdown Manager Turnover Reports, On-line/Off-line Daily Log, IE-A7 not met conditions, and events which result and Outage History was conducted to identify potential in a controlled shutdown leading to initiating events. Low power and non-power operation Associated SRs: a scram prior to reaching low- events were discussed as part of the system IE-A8 met at capability power conditions as specified in screening performed to resolve F&O Internal Event category I the standard. A review of historical (IE)-A5, discussed above.

IE-A9 met at capability events, plant operating history, and category I interviews with plant personnel are This SR is judged to now be met, based on also required by the standard. consideration of shutdown and low power events and unplanned shutdowns. Associated SRs IE-A8 and IE-A9 are also judged to be met at capability category II based on interviews having been conducted, and on review of operating history for precursor events.

A2-3

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level IE-C5 IE-C5-01 (Initiating F Closed Initiating event frequencies are An assessment was performed to determine whether event frequency based converted to events per calendar use of unit specific initiating event frequencies would on a reactor year basis) year by multiplying by the site have an impact on applications. The conclusion of critical hours per calendar year this assessment was that the difference in CDF and IE-C5 not met factor calculated from site LERF are negligible and would not impact the results operating experience, instead of a of any risk-informed applications.

unit-specific factor as required by the standard. This distinguishes differences in the plant units' operating experience.

IE-C10 IE-C1 0-01 (Combination F Closed Use of plant specific information, This F&O was resolved by additional review and of component failure including common cause failure model update if required. A summary review of the with the unavailability of (CCF) treatment, plant-specific initiating event fault trees indicates that plant-specific other components) data, repair times, and the information, including CCF treatment, plant-specific applicability of mitigating function data, repair times, -and the applicability of the IE-C10 met success criteria in the initiating mitigating function success criteria are currently used event fault tree was not evident. in the PRA model. A detailed review was performed and documented to confirm that all the required plant-specific information is included in the initiating event fault trees.

IE-C14 IE-C14-01 (Interfacing F Closed There is no documented A table listing the containment penetrations and systems loss-of-coolant systematic review of all disposition regarding their potential as an ISLOCA accident (ISLOCA) containment penetrations for pathway was developed. A set of screening criteria .

frequency) potentiaiiSLOCAs, including was developed consistent with the SR requirement.

identification of screened These criteria were used explicitly to screen each IE-C14 not met penetrations and the basis for potentiaiiSLOCA pathway. The unscreened ISLOCA screening, and relevant flow paths are consistent with what is modeled in surveillance test procedures and RISKMAN .

their impact on the potential for an ISLOCA. Also, impact of surveillance testing was added to the documentation. ------- -- --

A2-4

Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition I F&O SR Topic Status Finding Disposition Level IE-C15 IE-C15-01 (Uncertainty F Closed No discussion of uncertainty Parametric uncertainty for IE frequencies is given in associated with initiating parameters for initiating event fault the DCPP PRA documentation as Range Factors events) trees was identified. (Error Factors) for loss of cooling accident (LOCA) IEs and alpha/beta values for gamma distributions.

IE-C15 not met Associated SR: IE-C1 met A2-5

Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level IE-01 IE-01-01 F Closed The documentation is not written in References to PLG-0637 as the basis have been (Documentation) a manner that facilitates PRA taken out and information has been included in the applications, upgrades, and peer new calculation revisions for system notebooks, ID-01 not met review. The peer review team initiating event notebooks, event tree notebooks, and identified that the existing other PRA development documentation.

Associated SRs: documentation heavily references IE-02 not met the original DCPP PRA documents, especially PLG-0637.

IE-03 not met This makes it difficult to understand details of the model, AS-C1 not met difficult to confirm that the model addresses PRA requirements, and SY-C1 not met difficult to update and use it for PRA applications. This finding OA-E1 not met applies to other elements of the standard besides IE.

QU-F1 not met LE-G 1 not met IFPP-81 not met IFS0-8 1 not met IFSN-A5 met IFSN-81 not met IFQU-81 met A2-6

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition I

Level I IE-02 IE-02-01 F Closed The peer review team identified All identified initiating event documentation I (Documentation) specific examples of deficiencies in deficiencies were addressed in the most recent model the documentation of initiating update.

IE-02 not met events which need to be addressed, including specific Associated SRs: references missing, addressing IE-A3 met dual unit loss of instrument air as an initiating event, identification of IE-A10 met "freeze dates," identification of credited operator recovery actions, IE-83 met at capability details of uncertainty parameters category II and Bayesian updating of data, details of initiating event fault trees IE-C2 met (see IE-C1 0-01 ), and comparison to generic data sources.

IE-C3 met IE-C4 met IE-C8 met IE-C9 met IE-C10 met IE-C12 met IE-01 not met - - - - - - L__ -- L__

A2-7

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level AS-A11 AS-A 11-01 (Transfer s Closed This F&O is a suggestion that the The transfers between event trees is statically set by between event trees and event tree transfers would be more the initiators in RISKMAN . By looking at the initiator, it preserving easily followed if they were is clear how the event trees are link and the order that dependencies) explicitly given in the event trees. they transfer.

AS-A11 met AS-83 AS-83-01 F Closed There does not appear to be a A review of phenomenological conditions was (Phenomenological review of phenomenological performed for all of the initiating events in the DCPP conditions created by conditions created by each PRA. This review was documented in Calculation 1.1.

accident progressions) accident sequence; thus, there As a result of this review several changes were made may be non-safety related to the DCPP PRA model to correctly account for the AS-83 not met components that are affected by phenomenological conditions.

an accident sequence that were Associated SRs: not reviewed for the accident AS-83 not met impact on the functionality of the component.

SY-A18 met SY-A21met SY-A23 met SY-814 (met)

AS-87 AS-87-01 (Time-phased F Closed Time-phased dependencies were Documentation was reviewed and inconsistencies dependencies) found to be modeled in the were identified and corrected.

accident sequences (e.g., AC AS-87 met power recovery and DC battery depletion.) However, the documentation has inconsistencies that need to be resolved .

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Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level AS-C2 AS-C2-01 (Documenting F Closed The processes used to develop See AS-A11-01 and AS-87-01 for resolutions. I processes used to accident sequences are not develop accident sufficiently documented, as noted sequences) in F&Os AS-A11-01 and AS 01, which identify issues related to AS-C2 not met the documentation of the accident sequence analyses.

SC-A1 SC-A 1-01 (Definition of F Closed Two definitions of core damage The definition of core damage dependent on core damage) are used in the documentation. collapsed water level was removed from the The first definition, peak node documentation. Modular Accident Analysis Program SC-A 1 not met temperature >1800°F, is a valid (MAAP) runs were updated using the core damage success criterion, and meets the definition of> 1800°F peak fuel temperature.

Associated SR: SC-A2 definition in Section 1-2 of the not met standard. However, the second criterion of "the time until the water level is collapsed below the top of active fuel" is not a valid definition since the definition of core damage as written in Section 1-2 requires the consideration of uncovery and heat-up, and this definition does not consider heat-up.

A2-9

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level SC-A4 SC-A4-01 (Shared F Closed The identification of shared This F&O has been resolved by further evaluation; no systems between units) systems between the units and PRA model changes were required. A review of the how they are credited is not mitigating systems credited in the PRA model for dual-SC-A4 not met documented. For example, no unit initiators identified only the DFO transfer system discussion on the diesel fuel oil as a shared mitigation system not specifically (DFO) transfer system is provided, evaluated. Other shared systems were identified although it is a known shared correctly. The model correctly credits the DFO system system. This is significant to with consideration made that both units are impacted.

ensure that a shared system is not inadvertently credited for both units With this F&O resolved, SR SC-A4 is met.

simultaneously if the system does not have that capacity.

SC-A5 SC-A5-01 (Mission F Closed No discussion could be found that MAAP Calculations were reviewed and run past times) verified that each accident 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to verify that a safe stable state was sequence actually reached a safe achieved. Residual heat removal (RHR) entry SC-A5 not met stable state at the minimum conditions were also reviewed and verified for the specified mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. applicable accident sequences.

With this F&O resolved, along with additional F&O SC-A5-02 (see below), SR SC-A5 is met at capability category 11/111.

SC-A5 SC-A5-02 (Mission F Closed Several accident sequences were See response to F&O SC-A5-01 (above).

times) identified where RHR entry conditions were met prior to SC-A5 not met 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but RHR was not required for success in the accident sequence. If RHR is not questioned, then the end state may not be stable since heat removal via the SGs will be diminished as decay heat lowers, and RHR will be required to maintain temperatures long term.

A2-10

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level SC-83 SC-83-01 (LOCA break F Closed The current success criterion for This F&O was resolved by an update to initiating event sizes) LOCAs is based on plant frequencies. Additional analyses have been capabilities and system responses. performed and break sizes have been identified. The SC-83 not met The specific break sizes medium LOCA transition size was updated, and the associated with the transitions frequencies of LOCAs adjusted.

Associated SRs: between the LOCA definitions SC-8 1 met at capability have not been adequately justified Upon resolution of this F&O, and additional F&O SC-category II by specific thermal-hydraulic 83-02 (below), the SR SC-83 will be met.

evaluations.

IE-84 met IE-C1 met IE-C 13 met at capability category 1/11 L _ _ __ _ _ _

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level SC-83 SC-83-02 (ISLOCA F Closed The thermal-hydraulic analysis for This F&O was resolved by conducting additional sizes) ISLOCA referenced for the analyses to validate or revise the current ISLOCA success criteria validation is based break sizes and corresponding success criteria and SC-83 not met on an 8-inch break size, and not plant impacts. Documentation was updated to on a 2-inch break size. The use of properly identify and validate assumptions on impacts Associated SR: an 8-inch break size is to the RHR pumps.

SC-8 1 met at capability inappropriate because the required category II equipment and timing associated Upon resolution of this F&O, and additional F&O SC-with responding to a 2-inch break 83-01 (above), the SR SC-83 will be met.

would be significantly different than the required equipment and timing associated with an 8-inch break.

In addition, the RHR pumps are assumed to be unavailable based I on conservative assumptions related to the effects of the ISLOCA; more realistic assumptions should be applied.

SC-84 SC-84-01 (Define large F Closed The analysis code used to This F&O has been resolved by additional reviews; no break LOCAs) establish success criteria has model updates were required. The success criteria known limitations with respect to its from the design basis analysis are consistent with the SC-84 met modeling of large LOCAs. The PRA success criteria for large LOCAs.

limitations of the code are not summarized anywhere in the analyses, so it is not clear that the limitations of the code were considered when developing the success criteria.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition I F&O SR Topic Status Finding Disposition I Level SC-84 SC-84-02 (Anticipated F Closed The discussions associated with Documentation was updated to be consistent with the Transient Without Trip the ATWT scenarios and the model.

(ATWT) definition) success criteria for ATWT are not I

consistent in the documentation SC-84 met with regards to parameters relevant to A TWT events. The actual criteria for plant-specific ATWT conditions needs to be defined, justified, and evaluated for system response required to mitigate the ATWT.

SC-85 SC-85-01 (Crediting F Closed In the documentation of the The impact of not crediting feed and bleed for small PORVs for comparison of success criteria to LOCA scenarios was determined to be approximately depressurization when similar plants, one outlier was 1E-8/yr CD F. Although the risk benefit for this credit is AFW not available) noted in the success criteria for a not significant, it could contribute some risk benefit in small LOCA without AFW certain configurations, such as an AFW pump being SC-85 met available. This is assumed to inoperable. Therefore, the DCPP PRA model has result in core damage, but the use been updated to ensure that small LOCA scenarios of power-operated relief valves correctly credit the use of feed and bleed when (PORVs) to depressurize and appropriate.

cooldown is credited at similar plants. The basis for not crediting the use or PORVs is not documented, and discussions with plant PRA personnel did not identify any reason that the PORVs could not be credited at DCPP.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level SC-C2 SC-C2-01 (Unclear F Closed The process followed for Removed the collapsed water level definition of core process of developing developing the success criteria for damage and now use peak node temperature of success criteria) each accident scenario is not greater than 1800°F.

clearly documented. For example, SC-C2 not met there are two definitions of core Limitations of computer codes addressed in SC damage used, the basis for the 01. Impact of ATWT success criteria addressed in timing of human actions is not SC-84-02.

clear (two criteria used- but nothing showing why both are acceptable), the limitations of the software used for the success criteria is not documented, etc.

SC-C3 SC-C3-01 (Documenting F Closed A review of many of the PRA This F&O has been resolved by a documentation sources of uncertainty) elements identified that there was update. Each PRA element calculation has been not summarization of the sources reviewed and the assumptions and sources of SC-C3 not met of uncertainty or assumptions uncertainty have been documented.

associated with the individual PRA Associated SRs: element. With this F&O resolved, SC-C3 is met.

IE-03 not met SY-C3 not met A2-14

Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level SY-A4 SY-A4-01 (Walkdowns F Closed Neither plant walkdowns nor This F&O has been resolved by providing additional and interviews) interviews with knowledgeable evidence that confirms the system analyses were plant personnel were performed to correctly developed, refined and maintained to reflect SY-A4 not met confirm that the systems analysis the as-built and as-operated plant.

~

correctly reflects the as-built, as-operated plant. Based on the maturity of the system models and their ongoing application at the plant, it is judged unlikely that additional walkdowns or interviews would identify significant deficiencies requiring model updates, and that the current system models reasonably reflect the as-built/as-operated plant condition and configuration.

Therefore, resolution of this F&O would not impact the calculations of risk changes for the RI-ISI Program.

SY- SY-A11-01 (Failures to s Closed Failures to run in first hour (rather Failure to run during first hour is considered in the A11 run in first hour) than over the entire 24-hour model. These failure probabilities are incorporated mission time) were not addressed . into the basic event for failure to run and adequately by creating a new basic event. account for the impact on component reliability. The This could lead to model update F&O addresses the ease of model update given that issues. only one basic event exists for two failure modes.

SY-A16 SY-A16-01 (Modeling of F Closed No pre-initiator human failure Pre-initiators review was performed and pre-initiator pre-initiators) events (HFEs) are modeled in the HFEs were identified in G.1 Revision 2. Several AFW system model. Since AFW is miscalibration and misposition HFEs were added to SY-A 16 not met) a standby system, at least one pre- the PRA model.

initiator HFE (e.g., failure to restore Associated SR: HR- pump after maintenance or testing)

A1 not met is expected to be in the model.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level I I

SY-A20 SY-A20-01 F Closed Simultaneous unavailability of This F&O has been resolved by examination of the  !

(Simultaneous redundant safety-related maintenance schedules and update of documentation.

unavailability of equipment due to a planned redundant SSCs) activity is excluded from consideration , consistent with SY-A20 not met Technical Specification (TS) 3.0.3 restrictions. This assumption is I probably not appropriate for nonsafety-related equipment, whose unavailability is not restricted by a TS. An example of this is multiple instrument air compressors concurrently out of service. I SY-A23 SY-A23-01 (Consistent F Closed Consistent system/component Changed basic event naming convention for all AFW system model failure mode nomenclature is used top events nomenclature) in all system notebooks, except the AFW notebook.

SY-A23 met SY-83 SY-83-01 (CCF groups) F Closed No documentation was found for Documentation was revised for all systems to the CCF group definition for the specifically list the common cause failures that are SY-83 not met safety injection (SI) top event. For modeled.

other systems, CCF groups appear to generally be defined inside of RISKMAN files but not in the documentation . - - - - - - - - - - - - - - - - - - - - - - - -- - -

A2-16

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level SY-88 SY-88-01 (Spatial and F Closed No discussion of spatial and This F&O has been closed with no action taken .

environmental hazards environmental dependencies, or Documentation of the effects of room heatup is impacting multiple room heatup and dependence on available and references plant specific room heatup SSCs) heating, ventilation and air calculations. These results are not reiterated within conditioning (HVAC) could be the individual system notebooks but system modeling SY-88 not met found in the sampled system is consistent with the room heatup calculations.

notebooks. The peer review team Associated SR: SY-814 subsequently identified additional met documentation that was available to potentially address these gaps.

SY-810 SY-810-01 (Modeling of F Closed The treatment of permissives and PG&E performed a systematic evaluation of permissive and interlocks could not be located in modeling of permissives and interlocks in the Internal interlocks) the system notebooks. Events PRA (lEPRA) and the Fire PRA (FPRA) and documented in PRA Calculation 14-01 , Revision 1.

SY-810 not met The evaluation includes identification and modeling of (1) those systems that are required for initiation and actuation of a system, (2) the conditions needed for automatic actuation (e.g., low vessel water level),

and (3) control features (e.g., protection and control permissive, lock-out signals, and component interlocks that are required to complete actuation logic, as required in the SR of Section 2 of AMSE/ANS RA-SA-2009 Standard . Based on the results of the review, permissive and interlocks of the following structures, systems, and components (SSCs) are included in the Internal Events model:

8701/8702, 8982A/B,ang 9003A/B, 8804A/B.

A2-17

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level SY-815 SY-815-01 (Inter- F Closed Human actions that had the To address this F&O, the DCPP procedures were system operator potential to impact multiple trains reviewed to identify realignment and calibration dependencies) of a given system (miscalibration) activities for all systems and components including and actions from one system that any dependencies between activities and SY-815 not met could impact the function of components.

another system are not addressed .

As a result of this review, numerous pre-initiator HFEs were identified in standby systems and were quantified using the EPRI Human Reliability (HRA)

Calculator THERP module. Although pre-initiator dependency across trains was identified due to misposition and included in the DCPP HFEs, none of the HFEs involved miscalibration across systems or trains.

A2-18

Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level SY-C2 SY-C2-01 F Closed The peer review team identified This F&O has no impact on the RI-ISI Program.

(Documentation) specific examples of deficiencies in Updating the documentation to address specific the documentation of system examples of missing information would not impact the SY-C2 not met models which need to be calculations of risk changes for the RI-ISI Program.

addressed, including documenting However, all identified documentation issues were Associated SRs: assumptions, references, HVAC resolved during the latest DCPP PRA model update.

SY-A22 met at capability dependencies, success criteria category II and timing, and discussion of available inventories of air, power, SY-81 met and cooling to support the mission time.

SY-83 not met SY-86 met SY-87 met at capability category II SY-89 met SY-811 met HR-A1 HR-A 1-01 (Pre-initiator F Closed The identification of pre-initiator To address this F&O, DCPP procedures were events) HFEs based on whether the reviewed to identify realignment and calibration procedure or practice involves activities. This review was performed in order to be HR-A1 not met realignment or calibration should consistent with the ANS/ASME Standard supporting be performed before screening requirements HR-A 1 and HR-A2.

Associated SRs: processes are applied.

HR-A2 not met As a result of this review, additional pre-initiator HFEs were identified for inclusion into the PRA model and SY-A16 not met were quantified using the EPRI HRA Calculator THERP module. These new HFEs were incorporated into the PRA model.

A2-19

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O  !

SR Topic Status Finding Disposition Level HR-A3 HR-A3-01 (Pre-initiator F Closed Pre-initiator HRA screening criteria To address this F&O, all of the screening criteria were events) could remove restoration errors reviewed and revised as necessary to ensure that the prematurely. If a system or train is criteria applied specifically to the component being HR-A3 met automatically actuated following an operated/calibrated. The DCPP procedures were then event, then a restoration error of reviewed against the new criteria to identify manual valves in the flow path realignment and calibration activities.

could be missed. Examples include mispositioning of a valve in the standby component cooling water (CCW) pump train if it receives an automatic start signal on low header pressure and misposition of a valve in Sl pump train if the valve does not automatically open on an engineered safety features actuation system (ESFAS) signal.

HR-C3 HR-C3-01 F Closed The pre-initiator HRA To address this F&O, DCPP procedures were (Consideration of documentation discusses the reviewed to identify realignment and calibration miscalibration) reasons for not including common activities. This review was performed in order to be miscalibration, but the standard consistent with the ANS/ASME Standard supporting HR-C3 not met requires inclusion of such requirements HR-A1 and HR-A2.

miscalibration events.

As a result of this review, additional pre-initiator HFEs were identified for inclusion into the PRA model and were quantified using the EPRI HRA Calculator THERP module. These new HFEs were incorporated into the PRA model.

A2-20

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition SR Topic LF&OI Status Finding Disposition eve HR-03 HR-03-01 (Pre-initiator F Closed The detailed discussion of pre- A new section dealing with procedure and human-HFEs) initiator HFEs does not discuss the machine interface quality has been added to the quality of procedures, DCPP pre-initiator HRA documentation.

HR-03 met at capability administrative controls, or man- 1 category I machine interface (MMI) 1 requirements in performing the assessments.

HR-E1 HR-E1-01 (Crediting F Closed Operator actions associated with A review was performed to verify that no manual manual verification steps starting pumps or aligning valves recovery for failure of an automatic signal that could when automatic are not credited even when the be credited was missed. In order to avoid actuation failed) emergency operating procedures unnecessary complexity in the PRA model, the scope (EOP) specifically states "Verify" of the review was limited to risk significant basic 1 HR-E1 met pump started or "Verify" valve events. The risk significant basic events were open/closed . In the event the reviewed in conjunction with the EOPs to determine 1 Associated SR: SY- automatic signal fails to start the whether any additional manual recoveries of automatic '

A 17met pump or align the valve, credit signal failures could be found. No additional operator 1 should be taken for the operator actions were identified that could mitigate the failure of backing up the automatic signal. an automatic signal for risk significant components.

Therefore, no change to the OCPP PRA model is required .

A2-21

Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level HR-E3 HR-E3-01 (Consistent F Closed There is no discussion in the HRA Operator interviews were re-performed and interpretation of documentation on how the specific documented for each applicable operator action.

procedures) scenarios discussed in operator talk-throughs were selected, the HR-E3 met at capability questions posed to the operators, category I the entire sequence of procedures followed in the response to the accident sequence, etc. Actual I operator interview sheets are not included; only a summary of the I discussion is provided. Without having the basis for why the scenarios discussed were selected, it is not possible to ensure that the most risk-significant or important operator actions were discussed.

Additionally, without the operator Interview sheets it is not possible to verify what the operators/trainers said, and that the responses were taken in context.

HR-E4 HR-E4-01 (Confirming F Closed Talk-throughs performed with Simulator observations were performed to validate response models via Operations and Training personnel response models.

simulator observations do not address confirming that the or talk-throughs) response models (i.e. thermal-hydraulic analysis codes) used to HR-E4 met at capability support the PRA are realistic.

category I) Additionally, no documentation of the use of simulator observations to confirm the response models can be found. - - - - -

A2-22

Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-1 . Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level HR-G5 HR.:.G5-01 (Verification F Closed For some HFEs, no basis for the Operator interviews were re-performed and of the time estimates in required time to perform the action documented in for each applicable operator action .

HRA via observation of is provided. Response times were verified via interviews.

simulator or walk-throughs)

HR-G5 met at capability category II Associated SRs:

HR-E3 met at capability category I HR-E4 met at capability  !

category I HR-G6 HR-G6-01 (Combining F Closed Two HFEs appear to be essentially The two HEPs never appear in the same cutsets I identical HFEs) identical with the same human because of the mutually exclusive house event error probability (HEP). These two impacts used in the top event split fractions.

HR-G6 met should be combined into one HFE, Because they do not appear in the same cutsets, the since the use of both could dependency between two HFEs is immaterial. The adversely affect the HRA current model is adequate and no model changes dependence analysis and the are needed.

impact of the state of knowledge correlation in the quantified results. Documentation changes were made to clarify the diesel fuel oil modeling. RISKMAN data descriptions

-~- ----

were also updated to avoid confusion .

A2-23

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level HR-G7 HR-G7-01 (HFE F Closed The HFE dependency The HRA dependency analysis was updated. The dependencies) HR-G7 documentation does not list a set updated documentation clearly describes the operator not met of operator actions that were actions evaluated and how the dependencies were evaluated or how the dependence evaluated.

between actions is determined.

The process to identify and evaluate HFE dependencies does not seem to provide a thorough means for identifying and accounting for dependent human actions.

HR-H2 HR-H2-01 (Staffing level F Closed The staffing levels credited in the This F&O has been resolved by a documentation assumed in HRA) HRA include personnel not on-site update; no model changes were required . All HFEs 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 7 days a week, but are were reviewed and updated to reflect actual on-site HR-H2 met available via call-in -so they staffing levels. There were no impacts to the should not be credited for shorter probabilities of existing HFEs.

term responses. Additionally, minimum Operations staffing levels should be used when evaluating the post-initiator recovery actions.

HR-12 HR-12-01 s Closed The peer review team identified This F&O has been resolved by a documentation (Documentation) specific examples of deficiencies in update.

the documentation of HRA that need to be addressed, including

_normal vs. minimum staffing levels, use of multiple procedures, editorial corrections, and significant digits in the HEPs.

A2-24

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level HR-12 HR-12-02 (Estimation of s Closed A screening value is used for post- This F&O has been resolved by additional review; no HEPs) initiator event ZHEAS6 (Failure to model changes were required. A review confirmed close header cross tie valves, that event ZHEAS6 is not a significant HFE from a risk FCV-495 and FCV-496.) This HFE importance standpoint and use of a screening value is is used in many accident therefore consistent with the standard.

sequences, including ISLOCA accident scenarios. The number of these scenarios and their use in  !

ISLOCAs indicate that they are relatively significant events which should not use a screening value.

DA-C1 DA-C 1-01 (Use of the F Closed It is not evident that recognized This F&O has been resolved by additional review; no I latest industry sources are utilized for CCF and model changes were required. The generic source of documentation for sse off-site power recovery data. CCF data was not clearly identified in the failure rate, CCF, and documentation, but a review determined that all CCF offsite power recovery) data are from NUREG/CR-6928 which is a current recognized source. Offsite power recovery data DA-C 1 not met comes from NUREG/CR (INEEL/EXT-04-02326).

DA-C4 DA-C4-01 (Basis for F Closed A clear basis for the identification Detailed documentation of the basis for component identification of an event of events as failures has not been failure identification was added to the DCPP Data as a failure) developed. Also, no evidence was Analysis Notebook.

found that degraded states were DA-C4 not met distinguished as being applicable (or not) as failures.

Associated SR: DC-C3 not met DA-C5 DA-C5-01 (Documenting F Closed Documentation is inadequate to This finding is related to the documented evaluation of evaluation of failure confirm whether component failures occurring close in time when compiling plant events) failures occurring close in time are reliability data. The documentation was updated to separately counted . include reference to the Maintenance Rule DA-C5 not met methodology. A single example of such failures was identified and corrected.

A2-25

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1 . Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition _I F&O SR Topic Status Finding Disposition Level I DA-C6 DA-C6-0 1 (Removing F Closed Some post-maintenance tests Data analysis was reviewed and post-maintenance post-maintenance have been included in the testing demands were removed from the counts.

events from demand accounting of demands and Updates to the impacted failure probabilities in the counts) operating hours for plant-specific model were made.

data, which conflicts with the DA-C6 met standard.

DA- DA-C1 0-01 (Planned F Closed There was no discussion The documentation for plant-specific data was C10 coincident regarding counting of successful updated to account for any component sub elements unavailability) demands when components are which may have unique demand counts.

decomposed into sub elements.

DA-C 10 not met DA- DA-C14-01 (Planned F Closed No assessment of routine planned Examined the 12-week rolling maintenance outage C14 coincident unavailability) maintenance activities for multiple window matrix at DCPP and did not identify any component unavailabilities, or planned, repetitive activity which would cause DA-C 14 not met documentation that Maintenance coincident unavailability due to maintenance for Rule practices do not allow for redundant equipment (both intra-system and Associated SR: SY-A20 routine instances of multiple trains intersystem). Calculation or modeling of coincident not met or equipment being unavailable, maintenance unavailability was therefore were identified in the unnecessary.

documentation.

DA- DA-C16-01 (Disposition F Closed Plant specific LOOP events are not The finding is related to gaps in documentation of the C16 of plant-specific loss-of- identified in the documentation. disposition of plant-specific LOOP events used in offsite power (LOOP) determining the initiating event frequency. A review of events) the LOOP initiator frequency determined that plant-specific LOOP events are properly considered in the DA-C16 met - L_ _ _ - - - - - - - - -

_~ermination of initiating event freguency.

A2-26

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level DA-D4 DA-D4-01 (Tests and F Closed The peer review team identified The Bayesian updating is done using the RISKMAN check of data updates) specific examples of deficiencies in Data Module. Throughout the process, RISKMAN the documentation of data which shows the analyst a plot of the prior distribution, and a DA-D4 met at capability need to be addressed, related to plot of the prior distribution together with the posterior category 111111 Bayesian update data checks. distribution. RISKMAN also shows various stats for these distributions such as the mean, median, and Associated SR: DA-E1 range factor. This process helps the analyst not met determine if the update and the distributions are valid and make sense.

The Bayesian update checks for all failure rates and all initiating events were added as an attachment to the PRA Data Update Documentation . All distributions, including priors and posteriors, with their plots and statistics are stored in the RISKMAN files.

DA-D6 DA-D6-01 (Documenting F Closed NUREG/CR-5485 was used for This F&O has been resolved by a documentation method and references CCF methodology; however this is update to include the applicable reference to in data calculation) not listed as a reference or in NUREG/CR-5485 for the generic data source for CCF.

discussions in the calculation.

DA-06 met at capability category Ill DA-D8 OA-D8-01 (Documenting F Closed No documentation of analysis This F&O has been resolved with no action taken.

evaluation of design done on impact on data of design The evaluation of the potential impact to PRA data changes on impact on changes (such as recirculation due to DCNs are made as part of the design change data) sump screen design change process and documented during the design change notices (DCNs), or new charging process. On a routine basis as part of model DA-08 not met pump DCNs) could be found in the maintenance, all design changes since the last model data calculation. update are re-reviewed for impacts on the model.

Based on the documented evaluation of OCNs, SR DA-D8 is judged to be met at capability category II since plant data are used for significant basic events.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level DA-E2 DA-E2-01 F Closed Documentation does not facilitate The information provided in the backup documents (Documentation) review. Additional uncontrolled was accurate and review of these documents did not backup materials such as result in a finding that would impact the PRA model.

DA-E2 not met spreadsheets are required for a All PRA data analysis documentation was updated to traceable basis for plant data. include all information in a single calculation file Associated SR: DA-05 without external attachments or spreadsheets, met at capability including data calculation files.

category Ill QU-81 QU-81-01 (RISKMAN s Closed The peer review team This suggestion F&O has been resolved by a code limitations) recommended that the documentation update to include the RISKMAN code quantification document include a limitations. The limitations of the RISKMAN code do specific section that discusses not adversely impact its use in the RI-ISI Program.

RISKMAN code limitations.

QU-C2 QU-C2-01 (HFE F Closed Human action dependencies are Refer to F&O HR-G?-01. There is no requirement in dependency) not evaluated with a minimum the standard to use any minimum HEP for dependent default value of the HEP to prevent actions, only to account for such dependencies.

QU-C2 not met underestimating risk.

QU-04 QU-04-01 (Comparison F Closed The documentation includes a Resolved and documented by performing a more in-to other similar plants) comparison of results to other depth comparison with other Westinghouse 4-loop similar plants, but causes of plants.

QU-04 met at capability significant differences are not category I identified.

QU-E1 QU-E1-01 (Uncertainty) s Closed A review of generic sources of This suggestion F&O has been resolved by a uncertainty was performed; documentation update. The assumptions and however, this analysis would be uncertainties associated with each technical element improved by a review of plant- of different hazard groups are identified in the specific sources of uncertainty. documentation. As suggested in this F&O, these documents have been updated by systematically reviewing PRA development documents (e.g ., system notebooks, success criteria notebook, event-tree notebooks, etc.).

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition I F&O SR Topic Status Finding Disposition Level QU-F2 QU-F2-01 F Closed The peer review team identified Quantification documentation updated to include items 1 (Documentation) specific examples of deficiencies in listed in the Supporting Requirement.

the documentation of quantification QU-F2 not met which need to be addressed as specified in the standard.

Associated SR: QU-F1not met QU-F6 QU-F6-01 (Documenting F Closed There was no definition for Definition of significant sequences and basic event definition of significant) significant basic event located in importance added to the quantification documentation.

the documentation.

QU-F6 not met LE-C1 LE-C1-01 (Plant-specific s Closed Containment challenges in high This suggestion F&O was closed with no action taken.

level 2 model) level requirement LE-B must be The containment structural capability has been compared to the containment assessed and documented adequately.

LE-C1 met at capability structural capability analysis category I described in high level requirement LE-D.

LE-C2 LE-C2-01 (Modeling of F Closed The LERF analysis states that All SAMG procedures were reviewed. No additional operator actions there are no post-core damage human actions were identified either because they following the onset of operator actions available or were already credited as part of core damage core damage) credited. However, a review of mitigation or because the non-prescriptive nature of plant procedures identified that SAMG procedures did not lend themselves to HRA LE-C2 not met there are several severe accident techniques.

mitigation guidelines (SAMG) procedures available that do include post-core damage actions that need to be reviewed and credited as applicable.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level LE-C3 LE-C3-01 (Crediting s Closed No repair of equipment, other than The impact of not including repair of equipment is repair of SSCs in the potential restoration of AC conservative in that no credit is taken. Furthermore, significant LERF power following a loss of station the larger uncertainty involved in estimating equipment sequences) power (LOSP) event, is credited in repair likelihood, especially post-core damage, could the LERF analysis. The recovery skew the existing LERF results. Therefore, the impact LE-C3 met at capability of offsite power is only credited of not meeting capability category II is conservative.

category I pre-core damage, but could be considered for post-core damage The conservative treatment of not crediting repair or scenarios. recovery of equipment does not reduce the risk importance of the system screened-in for RI-ISI I program.

LE-C4 LE-C4-01 (Feasibility of s Closed The LERF model does not credit Excluding mitigating actions from the PRA results in a scrubbing) mitigating actions (e.g., isolate the conservative calculation of LERF. This conservative ruptured steam generator after treatment is acceptable for systems in the scope of the LE-C4 met at capability core damage, depressurize the RI-ISI Program, except for the containment spray (CS) category I RCS and terminate the leak, system. For CS, not crediting scrubbing mitigation recover containment integrity.) could underestimate the change in LERF. However, Additional fission product the frequency of core damage sequences that would scrubbing provided by the still have the CS system available is not significant in containment sprays is not credited. typical pressurized water reactor PRAs, and the Because it is assumed that all operation of CS therefore has limited impact on LERF.

early releases are large, it is implied that all SG tube rupture (SGTR) and ISLOCA core damage sequences remain un-scrubbed.

LE-C9 LE-C9-01 (Equipment s Closed No credit is taken for any This suggestion F&O does not adversely impact the survivability or human equipment survivability or human RI-ISI Program. Excluding mitigating actions or action under adverse actions under adverse conditions equipment from the PRA results in a conservative environments) or after containment failure. calculation of LERF. This conservative treatment is acceptable for the RI-ISI Program.

LE-C9 met at capability category I A2-30

Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition I

F&O SR Topic Status Finding Disposition Level LE-C13 LE-C13-01 (Realistic s Closed All core damage events involving This suggestion F&O does not adversely impact the containment bypass either a spontaneous SGTR, RI-ISI Program. Credit for scrubbing of fission analysis) pressure induced SGTR, or a products is addressed by F&O LE-C4-01 (above.)

thermally induced SGTR event, as Conservative treatment of ISLOCA and induced LE-C13 met at capability well as ISLOCA, were SGTR impacts results in a conservative estimate of category I conservatively assumed to lead to LERF. This conservative treatment is acceptable for a large early release. In addition, the RI-ISI Program .

fission product scrubbing provided by the CSs is not credited.

LE-07 LE-07-01 (Realistic F Closed There is no traceable basis for the A systematic evaluation of containment penetrations containment isolation list of containment isolation (CI) was performed and documented in PRA analysis) valves that are present in the Calculation E.8 Revision 8 and in a separate model and the systematic spreadsheet. A set of screening criteria was LE-07 met at capability disposition of all of the containment developed consistent with the requirement of this SR, category II penetrations that are not in the and consistent with large early release definition.

model. Each containment penetration is dispositioned explicitly using this set of screening criteria.

This F&O is closed and has no impact in RI-ISI application.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level LE-E2 LE-E2-01 Best N/A The discussion in PRA No disposition is required for this best practice F&O.

(Documentation) Practice documentation associated with the plant damage state (PDS) and containment event tree (CET) descriptions are very detailed, easy to follow, and address many i more potential damage states than I typically evaluated in a LERF I I

analysis. There is sufficient information in the tables and write-ups to understand when equipment is failed due to post core melt and/or post containment failure environments. Additionally, environmental/spatial impacts are addressed and the basis for I

equipment nonsurvivability is clearly delineated .

LE-E2 LE-E2-02 (Definition of F Closed No actual calculation verifying the As documented in response to F&O LE-07 -01, Cl LERF with 3-inch 3-inch containment break size analysis was re-performed based on greater than 2-opening) which constitutes a large release inch definition of the large early release path.

exists.

LE-E2 met LE-F2 LE-F2-01 (Review of F Closed The LERF results documentation The seal LOCA split fractions were confirmed to not LERF sequences for does notTeflect the latest LERF have changed since the level 2 analysis was reasonableness) cutsets. Additionally, the results performed, so there are no model updates required to include an out-of-date assumption address this issue.

LE-F2 met on RCP seal LOCA sizes which needs to be deleted and actual The latest update to the quantification documentation detailed results presented. includes LERF cutsets and insights.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-1. Diablo Canyon Internal Events PRA Peer Review F&Os and Disposition F&O SR Topic Status Finding Disposition Level LE-G3 LE-G3-01 (Documenting F Closed The relative contribution of The quantification documentation was updated to LERF calculations) contributors is not documented in include the contribution to LERF from initiating events the LERF calculation, and the as well as other requirements from this SR.

LE-G3 not met information in the quantification calculation does not reflect the latest results, and does not include all the types of contributions discussed in this supporting requirement.

LE-G5 LE-G5-01 (Limitations in F Closed The limitations in the various This F&O has no impact on the RI-ISI Program. The the LERF analysis) portions of the LERF analyses that DCPP PRA model includes a complete level 2 would impact applications are not detailed analysis. There are currently no general LE-G5 not met identified or discussed. limitations in the LERF analysis that would impact applications. The F&O is related to documentation of limitations in the LERF analysis. Documenting the limitations of the LERF analysis would not impact the calculations of risk changes for the RI-ISI Pro_gram.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17-048 Table A-2. Diablo Canyon Internal Flood PRA Peer Review F&Os and Disposition SR Topic F&O Status Finding Disposition Level IF SO- IFSO-A1-01 (Applicable s Closed Not all external flooding sources The internal flooding PRA was updated to address this A1 external sources) are identified in the F&O. Identification of potential flood sources include documentation, and walkdown in-leakage from other flood areas. Tank inventories information does not identify tank were identified.

inventories.

IF SO- IFSO-A6-01 (Spray F Closed The walkdown reports identify The internal flooding PRA was updated to address this A6 protection) equipment which is protected from F&O. Discussion of what constitutes spray protection the effects of spray; however, the was enhanced.

IFSO-A6 met documentation does not discuss what is specifically credited as spray protection and the limitations of that protection. This could result in future plant modifications which alter the plant configuration in a manner which impacts the spray protection without being recognized as an impact to the PRA.

IF SO- IFS0 s Closed Sources of epistemic uncertainty Internal flooding documentation was updated with 83 01 (Uncertainty) related to flood sources are not assessment of uncertainty.

explicitly discussed.

IFSN- IFSN-A3-0 1(Automatic F Closed Relevant automatic or operator The internal flooding PRA was updated to address this A3 and/or operator responses to flood events which F&O. For infinite flood sources, and large flood responses) could terminate or contain flood sources, auto and/or operator responses to terminate propagation are not identified in or contain a flood were added.

IFSN-A3 not met the documentation.

IFSN- IFSN-A4-01 (Capacity of s Closed Details on the capacity of floor Internal flooding documentation was updated. In A4 drains, berm, dikes, etc.) drains and sumps, and the impact general, credit for dikes, berms, and curbs is not taken of berms, dikes, and curbs are not to terminate or contain flood propagation. Curbs are IFSN-A4 not met discussed in the documentation. discussed as a means to estimate water height in local These features in general are not area where flood originates.

credited, and a more realistic evaluation could be performed.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-2. Diablo Canyon Internal Flood PRA Peer Review F&Os and Disposition I SR Topic F&O Status Finding Disposition I Level IFSN- IFSN-A6-01 (Spray s Closed No detailed evaluation of potential The internal flooding PRA was updated to address this !

A6 targets) spray targets based on the F&O. For spray, see resolution of IFSO-A6-01. The distance from the source with distance criteria for adverse spray impact from consideration of the maximum pressurized pipe and high-energy flood sources were potential spray elevation and added to the documentation and were applied for specific propagation paths has spray scenario development.

been made.

IFSN- IFSN-A7 -01 (Flooding s Closed For flooding effects to SSCs other The internal flooding PRA was updated to address this A7 impacts on SSCs) than submersion, the F&O. For spray impact, spray target component documentation does not describe screening and spray scenario development for the effects in a manner which is unscreened components was performed, see easily verifiable. resolution of IFSO-A6-01 and IFSN-A6-01. The affected equipment due to submergence (and spray) for unscreened scenarios are listed in the documentation.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-2. Diablo Canyon Internal Flood PRA Peer Review F&Os and Disposition SR Topic F&O Status Finding Disposition Level I

IFSN- IFSN-A8-01 (Drain line F Closed The potential for inter-area The Internal Flooding PRA Report was updated and  !

A8 and back flow paths) propagation through various documents the identification of propagation pathways flowpaths identified in the standard at DCPP. Due to the open layout design and IFSN-A8-01 met at are not identified in the numerous openings in different elevations of the capability category II documentation. auxiliary building and turbine buildings (e.g., open stairways and grate-covered floor openings), floods originating in one level are expected to propagate freely to the basement of the building. Other progagation pathways involving unsealed cable trays, conduit and pipe penetrations were also considered and documented in the internal flooding update.

IFSN- IFSN-A9-01 (Flood F Closed No calculations determine the Flood calculations were performed for selected areas A9 depth and propagation) flooding rates and the time to where bounding assumptions were too severe and equipment damage. more detailed analysis was required, including flood IFSN-A9 not met areas with limited drainage paths and large flood source capacities. The calculations consider flood rates, flood propagation through door gaps, opening between rooms and floor drains. The flooding depth (level rise) timing is evaluated in the updated internal flooding PRA report.

IFSN- IFSN-A10-01 (Size of F Closed Evaluations of the flooding The Internal Flooding PRA was updated to address A10 flood sources) scenarios do not include the this F&O. The size of infinite flood sources, circulating impact of emptying a source on water, auxiliary saltwater and firewater from the raw IFSN-A 10 met the flood depth in the areas, or the water reservoir, were included in the flood scenario propagation of infinite water development along with the flood area, source, flood sources without operator action to rate, sse damage, and operator actions.

isolate the flood.

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-2. Diablo Canyon Internal Flood PRA Peer Review F&Os and Disposition SR Topic F&O Status Finding Disposition Level IFSN- IFSN-A 11-01 (Multi-unit F Closed The impact of large flooding For the turbine building flood scenarios, ASW and A11 effects) sources in areas that could impact circulating water piping failure is assumed to cause a both units has not been dual unit trip. ASW and circulating water pipe breaks IFSN-A 11 not met considered . The potential for a in the intake structure causing dual unit trip are not large circulating water or auxiliary considered credible scenarios. In response to this saltwater (ASW) flood event on the F&O, pipe failures in auxiliary building flood areas that common turbine building and are shared between the two units are included in the intake structure resulting in a dual- flood initiator frequency count for both units (see unit shutdown was identified. Appendix G of Section 9, Revision 1 of the Internal Flooding PRA Report)

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Enclosure 2 Attachment 2 PG&E Letter DCL-17 -048 Table A-2. Diablo Canyon Internal Flood PRA Peer Review F&Os and Disposition  !

SR Topic F&O Status Finding Disposition Level I

IFSN- IFSN-A12-01 (Screening F Closed Flooding scenarios are screened The scenarios in the Internal Flooding PRA Report A12 of flood scenarios) or assumed not to propagate were reviewed. Additional propagation scenarios based on drains, curbs and previously screened in Revision 0 were identified and IFSN-A12 met barriers between rooms, and the scoped in with flood source capacity and propagation screening implicitly assumes that paths considered in characterization and quantification the leak is smaller than the drain of the flood scenarios. In addition, select HFEs were capacity and/or that the operators developed to model the flood isolation for large flood take action to reduce or stop the sources such as firewater from the raw water reservoir. 1 flow before water backs up into the Failure of these HFEs results in additional PRA room and fails additional equipment damage beyond the original source flood equipment or propagates beyond area, such as both RHR pumps being damaged the room. The propagation whenever the 54-foot pipe tunnel in the auxiliary screening does not look at building is flooded beyond its capacity volume.

accumulation on the area where the water is going and whether equipment in that area would be impacted due to flood or whether the flood could propagate beyond the second flood area to another area and damage equipment.

IFPP- IFPP-A5-01 (Walkdown F Closed The walkdown documentation has Walkdown documentation was updated to include A5 documents) missing information associated missing information for all flooding sources.

with the flooding sources.

IFPP-A5 met A2-38