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

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

ML17138B093
Person / Time
Site:	Diablo Canyon
Issue date:	05/18/2017
From:	Welsch J Pacific Gas & Electric Co
To:	Document Control Desk, 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: ML17138B093 (46)

Text

Pacific Gas and Electric Company James M. Welsch Diablo Canyon Power Plant Vice President, Nuclear Generation Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 80 5. 545.3242 Internal: 691.3242 May 18, 2017 Fax: 805.545.4884 PG&E Letter DCL-17 -048 U.S. Nuclear Regulatory Commission 10 CFR 50.55a ATIN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Power Plant Unit 1 and Unit 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

Reference:

1. PG&E Letter DCL-17 -028, "DCPP Units 1 and 2 ASME Section XI lnservice Inspection Program Plan- Fourth 10-Year Inspection Interval," dated April18, 2017 (ADAMS Accession No. ML17116A048)

Dear Commissioners and Staff:

In accordance with the provisions of 10 CFR 50.55a(z)(1 ), Pacific Gas and Electric Company (PG&E) requests Nuclear Regulatory Commission (NRC) approval to continue to use an alternative to the ASME Section XI Code examination requirements for inservice inspection (lSI) of Class 1 and Class 2 piping welds (Table IWB-2500-1, Examination Categories 8-F and 8-J, and Table IWC-2500-1, Examination Categories C-F-1 and C-F-2) for Diablo Canyon Power Plant (DCPP)

Units 1 and 2 for the fourth lSI interval. The proposed alternative request for DCPP Unit 1 is provided in Enclosure 1 and the proposed alternative request for DCPP Unit 2 is provided in Enclosure 2. As documented in the enclosures, the proposed alternative request provides an acceptable level of quality and safety as required by 10 CFR 50.55a(z)(1 ).

The lSI Program Plan (ISIPP) for the fourth 10-year inspection interval for DCPP Units 1 and 2 was submitted recently (Reference 1). The proposed risk-informed inservice inspection (RI-ISI) program for Class 1 and 2 pipe welds is an integral part of the ISIPP. The RI-ISI program plan was developed in accordance with the alternative methodology provided in Electric Power Research Institute (EPRI)

Topical Report (TR) 112657, Revision 8-A, "Revised Risk-Informed lnservice .

Inspection Evaluation Procedure." EPRI TR-112657, Revision 8, has been reviewed and accepted by the NRC. The NRC staff has found EPRI TR-112657, Revision 8, acceptable for referencing in licensing applications to the extent specified and under A member of the STARS Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek

Document Control Desk PG&E Letter DCL-17-048 May 18, 2017 Page 2 the limitations delineated in the report and the associated NRC Safety Evaluation Report, dated October 28, 1999.

PG&E had received NRC approval to use the alternative RI-ISI program for Class 1 and 2 pipe welds in the second 10-year lSI interval by letter dated November 8, 2001. The alternative program was again approved by the NRC for the third 10-year lSI interval for DCPP Units 1 and 2 by NRC letter dated January 16, 2013. PG&E implemented the alternative RI-ISI program for Class 1 and 2 pipe welds with required periodic updates as authorized through the end of the third lSI interval.

PG&E hereby requests NRC approval to use the alternative RI-ISI methodology and updated program again during the fourth 10-year lSI interval for DCPP Units 1 and 2. The fourth inspection interval began on May 7, 2015, for Unit 1 and March 13, 2016, for Unit 2.

PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this letter. If you have any questions or require additional information, please contact Mr. Hossein Hamzehee at (805) 545-4720.

Sincerely,

~- ~

Vice President, Nuclear Generation rntt/4231/50652488-7/8 Enclosures cc: Diablo Distribution cc/encl: Kriss M. Kennedy, NRC Region IV Administrator Christopher W. Newport, NRC Senior Resident Inspector Balwant K. Singal, NRC Senior Project Manager State of California, Pressure Vessel Unit A member of the STARS Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek

Enclosure 1 PG&E Letter DCL-17-048 Diablo Canyon Power Plant Unit 1 10 CFR 50.55a Request for Approval of Alternative - Rl-151-1

Enclosure 1 PG&'E Letter DCL-17 -048 Diablo Canyon Power Plant Unit 1 10 CFR 50.55a Request Number Rl-151-1 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 IW8-2500-1, Examination Categories 8-F* and 8-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 1 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 IW8-2500-1, Examination Categories 8-F* and 8-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 1. 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 B-F welds are included in the RI-ISI program for other damage mechanisms, Alloy 600182/182 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.

1

Enclosure 1 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 1 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 February 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. PG&E Letter 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 1 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), reactor coolant pump (RCP) seal injection check valves CVCS-1-8372A, B, and C, and CVCS-1-8367 A , B, and C were replaced. Multiple welds were deleted, added, or re-named 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.

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 (MOR) 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. However, the model of record 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, 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 3

Enclosure 1 PG&E Letter DCL-17 -048 DC01 DC02 Change in ConsequenceiD Rank Rank Consequence Rank 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 1, S6-50-3-WIB-222 was selected.

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

1. During the Unit 1 Nineteenth Refueling Outage, an ultrasonic examination found an indication on Class 1 weld WIB-378 . Further evaluation determined this was a 3/4-inch circumferential embedded flaw and therefore not due to an active degradation mechanism.

2. The DCPP PRA was updated to Model DC03 in July 2015. In Model DC03, the total CDF is 5.72E-05/yr and the total LERF is 5.81 E-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 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 ran kings:

Change in ConsequenceiD DC02 Rank DC03 Rank Consequence Rank ACC02A H M High to Medium ACC02B H M High to Medium 4

Enclosure 1 PG&E Letter DC L-17-048 Change in ConsequenceiD DC02 Rank DC03 Rank Consequence Rank 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-P EN 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 SI03B 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 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 4.33E-08 in CDF and 1.75E-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 1 Risk Impact Results

.1RiskcoF .1RiskLERF System w/POD w/o POD w/POD w/o POD ReS -1.44E-08 1.65E-09 -5.80E-09 6.68E-10 eves -6.79E-09 -4.00E-09 -2.74E-09 -1.62E-09 SIS -1.28E-08 -7.19E-09 -5.16E-09 -2.91 E-09 RHRS -9.13E-09 -4.26E-09 -3.69E-09 -1.72E-09 ess 3.01E-12 3.01E-12 3.01 E-13 3.01E-13 5

Enclosure 1 PG&E Letter DCL-17 -048 t\RiskcoF t\RiskLERF System w/POD w/o POD w/POD w/o POD RWST -2.61E-10 -2.61E-10 -1.05E-10 -1.05E-10 ccw O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO FWS 3.15E-13 5.15E-13 3.15E-14 5.15E-14 MSS  ?.OOE-14  ?.OOE-14  ?.OOE-15  ?.OOE-15 AFW 1.15E-13 1.55E-13 1.15E-14 1.55E-14 Total -4.33E-08 -1.41 E-08 -1.75E-08 -5.69E-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.

• 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.

6

Enclosure 1 PG&E Letter DC L-17-048 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 1 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 May 7, 2025.

7

Enclosure 1 Attachment 1 PG&E Letter DCL-17 -048 DCPP Unit 1 - Inspection Location Selection Comparison Between Previously Approved and Revised Rl-151 Prog_ram by Risk Categorv Previously Approved Updated Risk Failure Potential 1 Consequence Code (Third Interval) (Fourth Interval) 5ystem( l Rank Category Weld Weld Category Rank OMs Rank Rl-151 Other( 2 l Rl-151 Other(Z)

Count Count TASeS, 8(3) 5(3) 8(3) 5(3)

ReS 2 High High Medium 8-J TT ReS 2 High High TASeS Medium 8-J 9 4 9 4 8-F 1 0 1 0 ReS 2 High High TT Medium 8-J 13 0 13 0 TASeS, eves 2 High High Medium 8-J 4 3 4 2 TT eves 2 High High TT Medium 8-J 4 0 4 0 SIS 2 High High TT Medium 8-J 16 4 16 4 RHR 2 High High TASeS Medium e-F-1 12 3 12 3 Medium None Low 0(4) 0(4) 8(4) 0(4)

ReS 4 (2) High 8-F (High) (PWSee) (Medium) 8-F 21 2 13 0 ReS 4 Medium High None Low 8-J 273 35 281 31 eves 4 Medium High None Low 8-J 56 6 56 8 eves 4 Medium High None Low e-F-1 20 2 20 0 SIS 4 Medium High None Low 8-J 29 4 29 10 SIS 4 Medium High None Low e-F-1 60 6 125 6 RHR 4 Medium High None Low e-F-1 177 18 177 18 RWST 4 Medium High None Low e-F-1 47 5 116 12 eew 4 Medium High None Low e-F-2 13 2 13 2 Al-1

Enclosure 1 Attachment 1 PG&E Letter DCL-17 -048 DCPP Unit 1 - 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( 1 )

Rank Category Weld Other( 2 ) Weld Other( 2l Category Rank OMs Rank RI-ISI RI-ISI Count Count TASCS, eves 5a Medium Medium TT Medium 8-J 2 1 0 0 eves 5a Medium Medium TT Medium 8-J 2 0 0 0 SIS 5a Medium Medium IGSCC Medium 8-J 12 2 12 2 SIS 5a Medium Medium TASCS Medium C-F-1 4 0 4 1 RCS 6a Low Medium None Low 8-J 4 0 4 0 eves 6a Low Medium None Low 8-J 9 0 8 0 eves 6a Low Medium None Low C-F-1 677 0 0 0 SIS 6a Low Medium None Low 8-J 135 0 135 0 SIS 6a Low Medium None Low C-F-1 160 0 65 0 RHR 6a Low Medium None Low 8-J 18 0 18 0 RHR 6a Low Medium None Low C-F-1 96 0 96 0 css 6a Low Medium None Low C-F-1 72 0 72 0 RWST 6a Low Medium None Low C-F-1 69 0 4 0 TASCS, eves 6b Low Low TT Medium 8-J 0 0 2 0 eves 6b Low Low TT Medium 8-J 52 0 54 0 SIS 6b Low Low IGSCC Medium 8-J 7 0 7 0 AFW 6b Low Low TT Medium C-F-2 17 0 19 0 Low TASCS Medium FWS 6b (5b) Low C-F-2 29 0 34 0 (Medium) --

(FAG) (High)

Al-2

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

System( l Rank Category Weld Weld Category Rank OMs Rank RI-ISI Other( 2 l RI-ISI Other( 2l Count Count RCS 7a Low Low None Low B-J 17 0 17 0 B-J 2 0 3 0 eves 7a Low Low None Low C-F-1 0 0 750 0 SIS 7a Low Low None Low B-J 209 0 209 0 SIS 7a Low Low None Low C-F-1 8 0 34 0 css 7a Low Low None Low C-F-1 12 0 12 0 RWST 7a Low Low None Low C-F-1 4 0 0 0 MSS 7a Low Low None Low C-F-2 116 0 120 0 AFW 7a Low Low None Low C-F-2 132 0 129 0 Low None Low FWS 7a (5b)

(Medium)

Low (FAC) (High)

C-F-2 53 0 57 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 148 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 1 Attachment 2 PG&E Letter DCL-17 -048 Attachment 2 PRA Technical Adequacy for Rl-151 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 Level 1 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 MORis DC03A which was updated in 2016 and incorporates Westinghouse safe shutdown 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 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 .

A2-1

Enclosure 1 Attachment 2 PG&E Letter DCL-17 -048 A.3 Review of Modeling Uncertainties Table A4-2 in PRA Calculation C.10 Revision 7, "PRA Technical Adequacy," dated March 2016, provides a list of 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 was 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 1 Attachment 2 PG&E Letter DCL-17 -048 I

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 Closed There is no evidence in the This F&O has been resolved by additional reviews; no review of each system). 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-A? IE-A?-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-A? 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 screening power conditions as specified in performed to resolve F&O Internal Event (IE)-A5, IE-A8 met at capability the standard. A review of discussed above.

category I historical events, plant operating history, and interviews with plant This SR is judged to now be met, based on IE-A9 met at capability personnel are also required by the consideration of shutdown and low power events and category I standard. 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 1 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 event F Closed Initiating event frequencies are An assessment was performed to determine whether frequency based on a converted to events per calendar use of unit specific initiating event frequencies would reactor year basis) year by multiplying by the site have an impact on applications. The conclusion of this critical hours per calendar year assessment was that the difference in CDF and LERF IE-C5 not met factor calculated from site are negligible and would not impact the results of any operating experience, instead of a 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 model of component failure including common cause failure update if required. A summary review of the initiating with the unavailability of (CCF) treatment, plant-specific 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 mitigating IE-C10 met success criteria in the initiating function success criteria are currently used in the PRA event fault tree was not evident. 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) potential interfacing systems loss was developed consistent with the supporting SR. I of cooling accidents (ISLOCAs), These criteria were used explicitly to screen each IE-C14 not met including identification of screened potentiaiiSLOCA pathway. The unscreened ISLOCA penetrations and the basis for flow paths are consistent with what is modeled in screening, and relevant RISKMAN.

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

A2-4

Enclosure 1 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-C15 IE-C15-01 (Uncertainty F Closed No discussion of uncertainty Parametric uncertainty for IE frequencies is given in the associated with initiating parameters for initiating event fault DCP PRA documentation as Range Factors (Error events) trees was identified. Factors) for loss of cooling accident (LOCA) IEs and alpha/beta values for gamma distributions.

IE-C15 not met Associated SR:

IE-C1 met IE-01 IE-01 -01 F Closed The documentation is not written in References to PLG-0637 as the basis have been taken (Documentation) a manner that facilitates PRA out and information has been included in the new applications, upgrades, and peer calculation revisions for system notebooks, initiating ID-01 not met review. The peer review team event notebooks, event tree notebooks, and other PRA identified that the existing 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 DA-E1 not met applies to other elements of the standard besides I E.

QU-F1 not met LE-G 1 not met IFPP-81 not met IFS0-81 not met IFSN-A5 met IFSN-81 not met IFQU-81 met A2-5

Enclosure 1 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-02 IE-02-01 F Closed The peer review team identified All identified initiating event documentation deficiencies (Documentation) specific examples of deficiencies were addressed in the most recent model update.

in the documentation of initiating 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-B3 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-D 1 not met AS- AS-A 11-01 (Transfer s Closed This F&O is a suggestion that the The transfers between Event Trees are statically set by A11 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 linked and the order dependencies) explicitly given in the event trees. that they transfer.

AS-A11 met A2-6

Enclosure 1 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-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 were dependencies) found to be modeled in the 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.

AS-C2 AS-C2-01 (Documenting F Closed The processes used to develop See AS-A11-01 and AS-87-01 for resolutions.

processes used to accident sequences are not develop accident sufficiently documented, as noted sequences) in F&Os AS-A11-01 and AS-87 -01, which identify issues AS-C2 not met related to the documentation of the accident sequence analyses.

A2-7

Enclosure 1 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-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: definition in Section 1-2 of the SC-A2 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.

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 as discussion on the diesel fuel oil a shared mitigation system not specifically evaluated.

(DFO) transfer system is provided, Other shared systems were identified correctly. The although it is a known shared model correctly credits the DFO system with system. This is significant to consideration made that both units are impacted.

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

units simultaneously if the system does not have that capacity.

A2-8

Enclosure 1 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-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 achieved.

sequence actually reached a safe Residual heat removal (RHR) entry conditions were SC-A5 not met stable state at the minimum also reviewed and verified for the applicable accident specified mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 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 RH R entry conditions were met prior to 24 SC-A5 not met hours, 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.

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 performed capabilities and system and break sizes have been identified. The medium SC-83 not met responses. The specific break LOCA transition size was updated, and the frequencies sizes associated with the of LOCAs adjusted.

Associated SRs: transitions between the LOCA SC-81 met at capability definitions have not been Upon resolution of this F&O, and additional F&O category II adequately justified by specific SC-83-02 (below), the SR SC-83 will be met.

thermal-hydraulic evaluations.

IE-84 met IE-C1 met IE-C13 met at capability category 1/11 A2-9

Enclosure 1 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 SC-83 not met success criteria validation is based break sizes and corresponding success criteria and on an 8-inch break size, and not plant impacts. Documentation was updated to properly Associated SR: on a 2-inch break size. The use of identify and validate assumptions on impacts to the SC-B1 met at capability an 8-inch break size is RHR pumps.

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

2-inch break 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 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 from the design basis analysis are consistent with the SC-84 met its 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.

A2-10

Enclosure 1 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-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 consistent in the documentation SC-84 met with regards to parameters relevant to ATWT 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 CDF. 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 been result in core damage, but the use updated to ensure that small LOCA scenarios correctly of power-operated relief valves credit the use of feed and bleed when appropriate.

(PORVs) to depressurize and 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.

A2-11

Enclosure 1 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 damage used, the basis for the timing of human actions is not Limitations of computer codes addressed in clear (two criteria used - but SC-84-01. Impact of ATWT success criteria nothing showing why both are addressed in SC-84-02.

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 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.

A2-12

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

pre-initiator HFE (e.g., failure to Associated SR: restore pump after maintenance or HR-A1 not met testing) is expected to be in the model.

SY- SY-A20-01 F Closed Simultaneous unavailability of This F&O has been resolved by examination of the A20 (Simultaneous redundant safety-related

• maintenance schedules and update of documentation .

unavailability of equipment due to a planned redundant SSCs) activity is excluded from SY-A20 not met consideration, consistent with Technical Specification (TS) 3.0.3 restrictions. This assumption is 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.

SY- SY-A23-01 (Consistent F Closed Consistent system/component Changed basic event naming convention for all AFW A23 system model failure mode nomenclature is used top events nomenclature) in all system notebooks, except the AFW notebook.

SY-A23 met A2-13

Enclosure 1 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-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.

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 the conditioning (HVAC) could be individual system notebooks but system modeling is SY-88 not met found in the sampled system consistent with the room heatup calculations.

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

SY- SY-B 10-01 (Modeling of F Closed The treatment of permissives and PG&E performed a systematic evaluation of modeling 810 permissive and interlocks could not be located in of permissives and interlocks in the Internal Events interlocks) the system notebooks. 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 ievel), 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, and 9003A/B, 8804A/B.

A2-14

Enclosure 1 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- SY-815-01 (Inter-system F Closed Human actions that had the To address this F&O, the DCPP procedures were 815 operator dependencies) potential to impact multiple trains reviewed to identify realignment and calibration of a given system (miscalibration) activities for all systems and components including any SY-815 not met and actions from one system that dependencies between activities and components.

could impact the function of 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 Analysis (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.

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 Updating the documentation to address specific in 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 A2-15

Enclosure 1 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-A1 HR-A 1-01 (Pre-initiator F Closed The identification of pre-initiator To address this F&O, DCPP procedures were reviewed events) HFEs based on whether the to identify realignment and calibration activities. This procedure or practice involves review was performed in order to be consistent with the HR-A1 not met realignment or calibration should ANS/ASME Standard SRs HR-A 1 and HR-A2.

be performed before screening Associated SRs: processes are applied. As a result of this review, additional pre-initiator HFEs HR-A2 not met were identified for inclusion into the PRA model and were quantified using the EPRI HRA Calculator THERP SY-A16 not met module. These new HFEs were incorporated into the PRA model.

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 realignment manual valves in the flow path 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.

A2-16

Enclosure 1 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-C3 HR-C3-01 F Closed The pre-initiator HRA To address this F&O, OCPP procedures were reviewed (Consideration of documentation discusses the to identify realignment and calibration activities. This miscalibration) reasons for not including common review was performed in order to be consistent with the miscalibration, but the standard ANS/ASME Standard SRs HR-A 1 and HR-A2.

HR-C3 not met requires inclusion of such 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.

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 OCPP quality of procedures, pre-initiator HRA documentation.

HR-03 met at capability administrative controls, or man-category I machine interface (MMI) 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 be when automatic are not credited even when the credited was missed. In order to avoid unnecessary actuation failed) emergency operating procedures complexity in the PRA model, the scope of the review (EOP) specifically states, "Verify" was limited to risk significant basic events. The risk HR-E1 met pump started or "Verify" valve significant basic events were reviewed in conjunction open/closed. In the event the with the EOPs to determine whether any additional Associated S R: automatic signal fails to start the manual recoveries of automatic signal failures could be SY-A17 met pump or align the valve, credit found. No additional operator actions were identified should be taken for the operator that could mitigate the failure of an automatic signal for backing up the automatic signal. risk significant components. Therefore, no change to the OCPP PRA model is required.

A2-17

Enclosure 1 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 operator interview sheets are not included; only a summary of the 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.

A2-18

Enclosure 1 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 The two HEPs never appear in the same cutsets identical HFEs) essentially identical with the same because of the mutually exclusive house event human error probability (HEP). impacts used in the top event split fractions. Because HR-G6 met These two should be combined they do not appear in the same cutsets, the into one HFE, since the use of dependency between two HFEs is immaterial. The both could adversely affect the current model is adequate and no model changes are HRA dependence analysis and the 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.

A2-19

Enclosure 1 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) documentation does not list a set updated documentation clearly describes the operator HR-G7 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. I HR-H2 HR-H2-01 (Staffing level F Closed The staffing levels credited in the This F&O has been resolved by a documentation I

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 update.

in 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.

A2-20

Enclosure 1 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 that close header cross tie valves, event ZHEAS6 is not a significant HFE from a risk FCV-495 and FCV-496.) This importance standpoint and use of a screening value is H FE 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-C1-01 (Use of the F Closed It is not evident that recognized This F&O has been resolved by additional review; no 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 comes DA-C 1 not met 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 - --

tv A2-21

Enclosure 1 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-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.

DA-C6 DA-C6-01 (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 I 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-C14 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: routine instances of multiple trains intersystem). Calculation or modeling of coincident SY-A20 not met or equipment being unavailable, maintenance unavailability was therefore unnecessary.

were identified in the documentation.

DA- DA-C16-01 (Disposition F Closed Plant specific loss of offsite power The finding is related to gaps in documentation of the I C16 of plant-specific loss-of- (LOOP) events are not identified in disposition of plant-specific LOOP events used in offsite power (LOOP) the documentation. 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 determination of initiating event frequency.

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Enclosure 1 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 Data Module. Throughout the process, RISKMAN in 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 II/III Bayesian update data checks. distribution. RISKMAN also shows various stats for these distributions such as the mean , median, and Associated SR: range factor. This process helps the analyst determine DA-E1 not met 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 .

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Enclosure 1 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-08 DA-08-01 (Documenting F Closed No documentation of analysis This F&O has been resolved with no action taken. The evaluation of design done on impact on data of design evaluation of the potential impact to PRA data due to changes on impact on changes (such as recirculation DCNs are made as part of the design change process data) sump screen design change and documented during the design change process.

notices (DCNs), or new charging On a routine basis as part of model maintenance, all DA-08 not met pump DCNs) could be fo,und in the design changes since the last model update are re-data calculation. reviewed for impacts on the model.

Based on the documented evaluation of DCNs, SR DA-08 is judged to be met at capability category II since plant data are used for significant basic events.

DA-E2 DA-E2-0J F Closed Documentation does not facilitate The information provided in the backup documents was (Documentation) review. Additional uncontrolled accurate and review of these documents did not result backup materials such as in a finding that would impact the PRA model. All PRA DA-E2 not met spreadsheets are required for a data analysis documentation was updated to include all traceable basis for plant data. information in a single calculation file without external Associated SR: attachments or spreadsheets, including data calculation DA-D5 met at capability 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-G7-01. 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-D4 QU-D4-01 (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-D4 met at capability significant differences are not category I identified . - - - - - ---- ------ - - - - -- - ------ ----- --

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Enclosure 1 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 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 of improved by a review of plant- 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.).

QU-F2 QU-F2-01 F Closed The peer review team identified Quantification documentation updated to include items (Documentation) specific examples of deficiencies listed in the SR.

in the documentation of QU-F2 not met quantification which .need to be  !

addressed as specified in the Associated SR: standard.

QU-F1 not 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-8 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.

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Enclosure 1 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-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.

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 event, is credited in the repair likelihood, especially post-core damage, could LERF analysis. The recovery of skew the existing LERF results. Therefore, the impact LE-C3 met at capability offsite power is only credited pre- of not meeting capability category II is conservative.

category I 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 program.

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Enclosure 1 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-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 SG after core damage, treatment is acceptable for systems in the scope of the LE-C4 met at capability depressurize the reactor cooling RI-ISI Program, except for the containment spray (CS) category I system (RCS) and terminate the system. For CS, not crediting scrubbing mitigation leak, recover containment could underestimate the change in LERF. However, integrity). Additional fission the frequency of core damage sequences that would product 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 unscrubbed.

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

LE-C9 met at capability category I LE- LE-C13-01 (Realistic s Closed All core damage events involving This suggestion F&O does not adversely impact the Rl-C13 containment bypass either a spontaneous SGTR, lSI Program. Credit for scrubbing of fission products is analysis) pressure induced SGTR, or a addressed by F&O LE-C4-01 (above.) Conservative thermally induced SGTR event, as treatment of ISLOCA and induced SGTR impacts LE-C13 met at capability well as ISLOCA, were results in a conservative estimate of LERF. This category I conservatively assumed to lead to conservative treatment is acceptable for the RI-ISI a large early release. In addition, Program.

fission product scrubbing provided by the containment sprays is not credited.

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Enclosure 1 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-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 valves was performed and documented in PRA analysis) that are present in the model and Calculation E.8 Revision 8 and in a separate the systematic disposition of all of spreadsheet. A set of screening criteria was developed LE-07 met at capability the containment penetrations that consistent with the requirement of this SR, and category II are not in the model. consistent with large early release definition . 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.

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 and containment event tree descriptions are very detailed, easy to follow, and address many more potential damage states than typically evaluated in a LERF 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 equipment nonsurvivability is clearly delineated. -----

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Enclosure 1 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-E2-02 (Definition of F Closed No actual calculation verifying the As documented in response to F&O LE-D7 -01, LERF with 3-inch 3-inch containment break size containment isolation analysis was re-performed based opening) which constitutes a large release on greater than 2 inches definition of the large early exists. release path.

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 not reflect 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.

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 SR.

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 detailed would impact applications are not analysis. There are currently no general limitations in LE-G5 not met identified or discussed. 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 Program.

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Enclosure 1 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 IFSO-A1 IFSO-A1-01 (Applicable s Closed Not all external flooding sources are The internal flooding PRA was updated to external sources) identified in the documentation, and address this F&O. Identification of potential walkdown information does not flood sources include in-leakage from other identify tank inventories. flood areas. Tank inventories were identified.

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

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

explicitly discussed.

IFSN-A3 IFSN-A3-01 (Automatic F Closed Relevant automatic or operator The internal flooding PRA was updated to and/or operator responses) responses to flood events which could address this F&O. For infinite flood sources, terminate or contain flood propagation and large flood sources, auto and/or operator IFSN-A3 not met are not identified in the responses to terminate or contain a flood documentation. were added.

IFSN-A4 IFSN-A4-01 (Capacity of s Closed Details on the capacity of floor drains Internal flooding documentation was updated.

drains, berm, dikes, etc.) and sumps, and the impact of berms, In general, credit for dikes, berms, and curbs dikes, and curbs are not discussed in is not taken to terminate or contain flood IFSN-A4 not met the documentation. These features in propagation. Curbs are discussed as a general are not credited, and a more means to estimate water height in local area realistic evaluation could be where flood originates.

performed.

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Enclosure 1 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-A6 IFSN-A6-01 (Spray targets) s Closed No detailed evaluation of potential The internal flooding PRA was updated to spray targets based on the distance address this F&O. For spray, see resolution from the source with consideration of of IFSO-A6-01. The distance criteria for the maximum potential spray adverse spray impact from pressurized pipe elevation and specific propagation and high-energy flood sources were added to paths has been made. the documentation and were applied for spray scenario development.

IFSN-A7 IFSN-A7-01 (Flooding s Closed For flooding effects to SSCs other The internal flooding PRA was updated to impacts on SSCs) than submersion, the documentation address this F&O. For spray impact, spray I

does not describe the effects in a target component screening and spray manner which is easily verifiable. scenario development for unscreened components was performed, see 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.

IFSN-A8 IFSN-AS-01 (Drain line and F Closed The potential for inter-area The Internal Flooding PRA Report was back flow paths) propagation through various updated and documents the identification of flowpaths identified in the standard propagation pathways at DCPP. Due to the IFSN-AS-01 met at capability are not identified in the open layout design and numerous openings category II documentation. in different elevations of the 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 UQdate.

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Enclosure 1 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-A9 IFSN-A9-01 (Flood depth F Closed No calculations determine the Flood calculations were performed for and propagation) flooding rates and the time to selected areas where bounding assumptions equipment damage. were too severe and more detailed analysis IFSN-A9 not met was required, including flood 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-A 10-01 (Size of flood F Closed Evaluations of the flooding scenarios The internal flooding PRA was updated to A10 sources) do not include the impact of emptying address this F&O. The size of infinite flood a source on the flood depth in the sources, circulating water, auxiliary saltwater IFSN-A10 met areas, or the propagation of infinite (ASW), and firewater from the raw water water sources without operator action reservoir were included in the flood scenario to isolate the flood. development along with the flood area, source, flood rate, sse damage, and operator actions. - - - ----- - -

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Enclosure 1 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 sources For the turbine building flood scenarios, ASW A11 effects) in areas that could impact both units and circulating water piping failure is assumed has not been considered. The to cause a dual unit trip. ASW and circulating IFSN-A11 not met potential for a large circulating water water pipe breaks in the intake structure or ASW flood event on the common causing dual unit trip are not considered turbine building and intake structure credible scenarios. In response to this F&O, resulting in a dual-unit shutdown was pipe failures in auxiliary building flood areas identified. that are shared between the two units are included in the flood initiator frequency count for both units (see Appendix G of Section 9, Revision 1 of the Internal Flooding PRA Report).

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

second flood area to another area and damage equipment.

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

IFPP-A5 met --

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