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 Requirement for Class 1 and 2 Piping Welds

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Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Examination Requirement for Class 1 and 2 Piping Welds
ML12025A084
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 01/20/2012
From: Becker J
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-12-007, OL-DPR-80, OL-DPR-82
Download: ML12025A084 (28)
v • d • e


Text

PacificGas and ElectricCompany' James R.Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 January 20, 2012 805.545.3462 Internal: 691.3462 Fax: 805.545.6445 PG&E Letter DCL-12-007 U.S. Nuclear Regulatory Commission 10 CFR 50.55a ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 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

Dear Commissioners and Staff:

In accordance with the provisions of 10 CFR 50.55a(a)(3)(i), Pacific Gas and Electric Company (PG&E) requests NRC approval to use an alternative to the ASME Section XI Code examination requirements for inservice inspection of Class 1 and 2 piping welds (Categories B-J, C-F-i, and C-F-2) for Diablo Canyon Power Plant (DCPP) Units 1 and 2. The proposed alternative, "10 CFR 50.55a Request Number RI-ISI-INT3," provides an acceptable level of quality and safety as required by 10 CFR 50.55a(a)(3)(i).

The proposed risk-informed inservice inspection (RI-ISI) program for Class 1 and 2 pipe welds is an integral part of the DCPP Units 1 and 2 Inservice Inspection Program Plan (ISIPP) for the third inspection interval. The RI-ISI plan was developed in accordance with the methodology provided in Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure." EPRI TR-1 12657, Revision B, has been reviewed and accepted by the Nuclear Regulatory Commission (NRC). The NRC staff has found EPRI TR-1 12657, Revision B, acceptable for referencing in licensing applications to the extent specified and under the limitations delineated in the report and the NRC Safety Evaluation Report, dated October 28, 1999.

PG&E had received NRC approval to use the RI-ISI program for Class 1 and 2 pipe welds, by letter dated November 8, 2001. The relief was authorized for the second ten-year ISI interval for DCPP Units 1 and 2. PG&E continued to use the RI-ISI program for Class 1 and 2 pipe welds, approved for the second DCPP ISI interval in the second period of that interval, through the first period of the third interval, A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Cattaway . Comanche Peak . Diablo Canyon . Palo Verde . San Onofre . South Texas Project . Wolf Creek

Document Control Desk PG&E Letter DCL-12-007 January 20, 2012

'8 Page 2 completing the full ten years of RI-ISI program application. Currently, PG&E is using the RI-ISI program for the ongoing second period of the third interval. PG&E has identified that such use exceeded the original approval time endpoint (i.e., the end of the second interval), and that issue is documented in the plant corrective action program.

PG&E requests NRC approval to use the alternative updated RI-ISI program during the third ten-year ISI interval for DCPP Units 1 and 2. The third inspection interval began January 1, 2006, for Unit 1 and July 1, 2006, for Unit 2.

Pending the receipt of NRC approval, PG&E proposes to continue using the alternative updated RI-ISI program discussed here during the ongoing second-period of the third inspection interval, including the DCPP Unit I seventeenth refueling outage (1R17), currently scheduled to begin April 22, 2012, and requests NRC approval by that date.

PG&E intends to incorporate this risk-based program for Class 1 and 2 piping weld inspection into the DCPP ISIPP - third ten-year inspection interval, as shown in the enclosure.

There are no new or revised regulatory commitments as defined by the Nuclear Energy Institute 99-04, "Guidelines for Managing NRC Commitment Changes," dated July 1999, in this submittal.

If you have any questions regarding the information enclosed, or other ISI program activities, please contact Mr. Patrick Nugent, Manager, Technical Support Engineering, at (805) 545-4701.

James R. Becker Site Vice President rntt/4231/50441474 Enclosure cc: Diablo Distribution cc/encl: Elmo E. Collins, NRC Region IV Michael S. Peck, NRC Senior Resident Alan B. Wang, NRR Project Manager State of California, Pressure Vessel Unit A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway - Comanche Peak - Diablo Canyon

  • Palo Verde - San Onofre - South Texas Project - Wolf Creek

Enclosure PG&E Letter DCL-12-007 Diablo Canyon Power Plant Inservice Inspection Program Third 10-Year Inspection Interval Revision 1 10 CFR 50.55a Request Number RI-ISI-INT3-U1 & 2 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

Diablo Canyon Power Plant Inservice Inspection Program Plan Third 10-Year Inspection Interval Revision 1 10CFR50.55a Request Number RI-ISI-INT3-UI&2 Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

Unit 1 USNRC Docket: 50-275 Facility Operating License: DPR-80 Commercial Operation Date: May 7, 1985 Third Interval Start Date: January 1, 2006 Unit 2 USNRC Docket: 50-323 Facility Operating License: DPR-82 Commercial Operation Date: March 13, 1986 Third Interval Start Date: July 1, 2006 Approval Record Prepared by: _ > t/ ,

PRA Independent Te cal Review: _/,

ISI Independent Technical Review: O q-i^ --

Approved by:

10CFR50.55a Request Number RI-ISI-INT3-Ul&2 Proposed Alternative In Accordance with 10CFR50.55a(a) (3) (i)

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

II. Applicable Code and Edition The Diablo Canyon Power Plant (DCPP) Unit 1 and 2 ISI program is based on the 2001 Edition of ASME Section XI through the 2003 Addenda.

III. Applicable Code Requirement Table IWB-2500-1, Examination Category B-F* and Category B-J Table IWC-2500-1, Examination Category C-F-I and Category C-F-2 IV. Reason For Request The continued use of a risk 7 informed process as an alternative for the selection of Class 1 and Class 2 Piping Welds for examination is requested for the Third Interval of Units 1 and 2.

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 and Unit 2 ISI program for the examination of Class 1 and Class 2 piping welds is proposed in accordance with a risk-informed process developed based on EPRI TR-112657, Revision B-A, with identified differences, and with additional guidance taken from ASME Code Case N-578. A request to utilize this process for the second inspection interval was submitted on 02/16/2001. The NRC approved this request on 11/08/2001 (TAC No. MB1203 - Unit 1, and TAC No. MB1204 - Unit 2). The RI-ISI plan approved for the second interval detailed 10 years of examinations; those assigned to the first inspection period were

performed in the first period of. the third interval. PG&E identified that although this use of the RI-ISI program schedule completed the full 10 years of applicability, it exceeded the approval time endpoint for application. A corrective action program document has been created to identify and resolve this issue. This request for alternative is being submitted in the second period of the third interval and is for the third interval only. On approval, and including examinations performed to date, PG&E intends to complete the full 10-year RI-ISI examinations as scheduled during the third interval.

In the original second interval submittal, DCPP committed to review and adjust the risk ranking of piping segments as a minimum on an ASME period basis. The first period of implementation of the RI-ISI program was the second period of Interval 2, which ended August, 2002 (Unit 1) and March, 2003 (Unit 2). To satisfy the periodic review requirements, an evaluation and update was performed in accordance with the Nuclear Energy Institute document 04-05, "Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems", published in April, 2004.

In accordance with NEI 04-05, the following aspects were considered during the review:

  • Plant Examination Results 0 Piping Failures

-Plant Specific Failures

-Industry Failures

  • Plant Design Changes

-Physical Changes

-Programmatic Changes

-Procedural Changes

  • Changes in Postulated Conditions

-Physical Conditions

-Programmatic Conditions The updated program resulting from this review is the subject of this proposed alternative.

  • Note that although Category B-F welds are included in the RI-ISI program for othtr damage mechanisms, Alloy 600/82/182 examinations are conducted per Code Cases N-722-1 and N-770-1.

unit 1 In accordance with the guidance provided by NEI 04-05, a table is provided as Attachment 1-1 identifying the number of welds added to and deleted from the originally approved RI-ISI program for the second interval for Unit 1. The additions to the original program are attributable to 5 specific action(s):

1) The CVCS system has 6 additional welds. These additional welds are due either to newly identified welds, or adding reducers to valves that were replaced where the original valve was welded directly to piping.
2) The SIS system has 2 additional welds due to the addition of a void header system.
3) The SIS system has 4 welds deleted due to a walkdown of the system indicating these welds did not exist.
4) The FW system has 4 additional welds. These additional welds are due either to newly identified welds, or welds added during replacement of piping due to FAC wall thinning.
5) The change to the 2001 edition of ASME Section XI with 2003 Addenda sees the inclusion of Auxiliary Feedwater (AFW) resulting in the addition of 148 welds.

In Revision DC01 to the DCPP PRA Consequence Case MFW-I changed consequence rank from Medium to Low. As a result of this change, sixteen Consequence Segments changed from a risk rank of Medium to a risk rank of Low. Four FW Risk segments changed from Cat 5 to Cat 6, eight FW Risk segments changed from Cat 6 to Cat 7, and thirty four MS segments changed from Cat 6 to Cat 7.

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 a reduction of -9.89E-08 in regards to CDF and -9.89E-10 in regards to LERF (Reference 14, Risk Impact Analysis for DCPP 12P3).

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 10CFR50.55a(a) (3) (i) it is requested that the proposed alternative be authorized.

Unit 2 In accordance with the guidance provided by NEI 04-05, a table is provided as Attachment 1-2 identifying the number of welds added to and deleted from the originally approved RI-ISI program for the second interval for Unit 2. The additions to the original program are attributable to four specific action(s):

1) The CVCS system has 4 additional welds, these additional welds are due to newly identified welds.
2) The SIS system has 2 additional welds due to the addition of a void header system.
3) The SIS system has 4 welds deleted due to a walkdown of the system indicating these welds did not exist.
4) The change to the 2001 edition of ASME Section XI with 2003 Addenda sees the inclusion of Auxiliary Feedwater (AFW) resulting in the addition of 145 welds.

In Revision DC01 to the DCPP PRA Consequence Case MFW-I changed consequence rank from Medium to Low. As a result of this change, sixteen Consequence Segments changed from a risk rank of Medium to a risk rank of Low. Four FW Risk segments changed from Cat 5 to Cat 6, eight FW Risk segments changed from Cat 6 to Cat 7, and thirty four MS Risk segments changed from Cat 6 to Cat 7.

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 a reduction of -1.22E-07 in regards to CDF and -1.22E-09 in regards to LERF (Reference 14, Risk Impact Analysis for DCPP 12P3).

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 10CFR50.55a(a) (3) (i) it is requested that the proposed alternative be authorized.

VI. PRA Quality PRA Quality assessment is provided in Attachment 2

VII. Duration of Proposed Alternative This alternative will be used for DCPP Units 1 and 2 until the end of each unit's third ten-year ISI program inspection interval, subject to the review and update guidance of NEI 04-

05. The third inspection interval is currently scheduled to end on or about May 6, 2015 (Unit 1) and March 12, 2016 (Unit 2).

This proposed alternative addresses 3 inservice inspection periods (nominally 40 months each) , comprising one inspection interval. As a result of initial implementation of the RI-ISI Program in the second period of the second interval, this RI-ISI program includes the examinations conducted in the first inspection period of this third interval, essentially restating that portion of the program previously approved during the second interval, updated as noted above. Therefore by the end of this third interval, PG&E intends to have performed the full 10 years of required examinations detailed in the RI-ISI program.

Attachment 1-1 DCPP Unit 1 - Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Failure Original Updated Consequence Potential Code System(1)

Category Rank Rank DMs Rank Category Weld Other(2) RI-ISI OtherW2)

Count Count 2 High High TASCS, Medium B-J 8 3(3) 8 5(3)

RCS TT RCS 2 High High TASCS Medium B-J 9 3 9 4 B-F 1 0 1 0 RCS 2 High High TT Medium B-J 13 2 13 0 CVCS 2 High High TASCS, Medium B-J 4 3 4 3 TT _ _ _ _ ____

CVCS 2 High High TT Medium B-J 4 1 4 0 SIS 2 High High TT Medium B-J 16 4 16 4 RHR 2 High High TASCS Medium C-F-i 12 3 12 3 B-F 21 2 21 2 RCS 4 Medium High None Low B-J 273 35 273 35 CVCS 4 Medium High None Low B-J 56 6 56 6 CVCS 4 Medium High None Low C-F-i 20 2 20 2 SIS 4 Medium High None Low B-J 29 4 29 4 SIS 4 Medium High None Low C-F-I 60 6 60 6 RHR 4 Medium High None Low C-F-i 177 18 177 18 RWST 4 Medium High None Low C-F-i 47 5 47 5 CCW 4 Medium High None Low C-F-2 13 2 13 2 FWS 5a (3) Medium Medium TASCS Medium C-F-2 28 3 Changed to (High) (FAC) (High) FWS 6b (5)

CVCS 5a ,TT Medium Medium TASCS, I Medium B-J 2 1 2 1

Attachment 1-1 (Cont'd)

DCPP Unit 1 - Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Failure Original Updated System Consequence Potential Code Category Rank DMs Rank CategoryRI-ISI Other (2) Weld RI-ISI Other(2 Count Count CVCS 5a Medium Medium TT Medium B-J 2 0 2 0 SIS 5a Medium Medium IGSCC Medium B-J 12 2 12 2 SIS 5a Medium Medium TASCS Medium C-F-I 4 0 4 0 Low None Low C-F-2 50 0 Changed Cagdt to FWS 6a (3) Lw Medium Noe Lw (High) (FAC) (High) FWS 7a (5)

RCS 6a Low Medium None Low B-J 4 0 4 0 CVCS .6a Low Medium None Low B-J 9 0 9 0 CVCS 6a Low Medium None Low C-F-I 671 0 677 0 SIS 6a Low Medium None Low B-J 135 0 135 0 SIS 6a Low Medium None Low C-F-I 158 0 158 0 RHR 6a Low Medium None Low B-J 18 0 18 0 RHR 6a Low Medium None Low C-F-I 96 0 96 0 CSS 6a Low Medium None Low C-F-I 72 0 72 0 RWST 6a Low Medium None Low C-F-I 69 0 69 0 MSS 6a Low Medium None Low C-F-2 116 0 Changed to MSS 7a FWS 6b (5) Low Low TASCS Medium C-F-2 Previously 29 0 (Medium) (FAC) (High) FWS 5a (3)

CVCS 6b Low Low TT Medium B-J 52 0 52 0 SIS 6b Low Low IGSCC Medium B-J 7 0 7 0 AFW 6b Low Low None Low C-F-2 0(4) 0 4) 16 0 Lo oe LwPreviously 5 FWS 7a (5) Low None Low C-F-2 FWS 54 0 (Medium) (FAC) (High) FWS 6a (3)

MSS 7a Low Low None Low C-F-2 Previously MSS 6a 116 0

Attachment 1-1 (Cont'd)

DCPP Unit 1 - Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Failure Original Updated 1 Rank Consequence Potential Code Category Weld Weld SystemC' 2)

Category Rank DMs Rank Count RI-ISI Other(2) Count RI-ISI Other(2)

RCS 7a Low Low None Low B-J 17 0 17 0 CVCS 7a Low Low None Low B-J 2 0 2 0 SIS 7a Low Low None Low B-J 209 0 209 0 SIS 7a Low Low None Low C-F-1 8 0 6 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 4 0 AFW 7a Low Low None Low C-F-2 0 (4) 0 (4) 132 0 Notes

1. Systems were described in Table 3.1-2 of the original submittal, with the exception of AFW - Auxiliary Feedwater System, as follows: RCS - Reactor Coolant System; CVCS - Chemical and Volume Control System; SIS -

Safety Injection System; RHR - Residual Heat Removal System; CSS - Containment Spray System; CCW - Component Cooling Water System; MSS - Main Steam System; FWS - Main Feedwater System; RWST - Refueling Water Storage Tank. The ASME Code Class 2 AFW system consists of 8 segments with 149 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. 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. 1 of the 3 welds selected for RI-ISI is the surge line elbow and is not counted as part of the weld count.
4. Due to a change in ASME Section XI Code criteria, 3" NPS Class 2 auxiliary feedwater piping was added to the ISI Program, and therefore the RI-ISI Program, for the first time during the Third ISI interval. As such, there were no welds associated with this piping during the original RI-ISI application.

Attachment 1-2 DCPP Unit 2 - Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Failure Original Updated system", Sytn~~Rank Consequence Potential Category Code Weld WeldOte(2 Category Rank DMs Rank Count RI-ISI Other(2) RI-ISI OtherC(2) 1 CountCount RCS 2 High High TASCS, Medium B-J 9 3(3) 9 6(3)

TT RCS 2 High High TASCS Medium B-J 10 4 10 4 B-F 1 0 1 0 RCS 2 High High TT Medium B-J 13 3 13 0 CVCS 2 High High TASCS, TT Medium B-J 5 3 5 3 CVCS 2 High High TT Medium B-J 3 1 3 1 SIS 2 High High TT Medium B-J 18 6 18 6 RHR 2 High High TASCS Medium C-F-i 11 3 11 3 B-F 21 2 21 2 RCS 4 Medium High None Low B-J 277 34 277 34 CVCS 4 Medium High None Low B-J 92 11 92 11 CVCS 4 Medium High None Low C-F-i 21 2 21 2 SIS 4 Medium High None Low B-J 30 4 30 4 SIS 4 Medium High None Low C-F-i 68 7 68 7 RHR 4 Medium High None Low C-F-i 175 18 175 18 RWST 4 Medium High None Low C-F-i 45 5 45 5 CCW 4 Medium High None Low C-F-2 12 2 12 2 5a (3) Medium Medium TASCS Medium C-F-2 28 3 Changed to FWS (High) (FAC) (High) FWS 6b (5)

CVCS 5a Medium Medium TASCS, Medium B-J 2 1 2 1 11 1 1 ~TT IIII

Attachment 1-2 (Cont'd)

DCPP Unit 2 - Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Failure original Updated System(1) Consequence Potential Code Systemgoy el Wl Category Rank Rank DMs Rank Category Weld RI-ISI Other(2) Weld RI-ISI Other (2)

Count Count CVCS 5a Medium Medium TT Medium B-J 2 0 2 0 SIS 5a Medium Medium IGSCC Medium B-J 13 2 13 2 SIS Sa Medium Medium TASCS Medium C-F-I 4 0 4 0 FWS 6a (3) Low Medium None Low C-F-2 37 0 Changed to (High) (FAC) (High) FWS 7a (5)

RCS 6a Low Medium None Low B-J 3 0 3 0 CVCS 6a Low Medium None Low B-J 8 0 8 0 CVCS 6a Low Medium None Low C-F-I 667 0 671 0 SIS 6a Low Medium None Low B-J 134 0 134 0 SIS 6a Low Medium None Low C-F-I 158 0 158 0 RHR 6a Low Medium None Low B-J 20 0 20 0 RHR 6a Low Medium None Low C-F-I 85 0 85 0 CSS 6a Low Medium None Low C-F-i 72 0 72 0 RWST 6a Low Medium None Low C-F-I 72 0 72 0 MSS 6a Low Medium None Low C-F-2 118 0 Changed to MSS 7a FWS 6b (5) Low Low TASCS Medium C-F-2 Previously 28 0 (Medium) (FAC) (High) FWS 5a (3)

CVCS 6b Low Low TT Medium B-J 52 0 52 0 SIS 6b Low Low IGSCC Medium B-J 7 0 7 0 AFW 6b Low Low None Low C-F-2 0(4) 0(4) 15 0 FWS 7a (5) Low Low None Low C-F-2 Previously 37 0 (Medium) (FAC) (High) FWS 6a (3)

MSS 7a Low Low None Low C-F-2 Previously -118 0 MSS 6a

Attachment 1-2 (Cont'd)

DCPP Unit 2 - Inspection Location Selection Comparison Between Original Approved and Revised RI-ISI Program by Risk Category Risk Failure Original Updated System(1) Consequence Potential Code Ste1 Rank Category Weld , Weld Category Rank DMs Rank RI-ISI Other(2) RI-ISI Other(2)

Count Count RCS 7a Low Low None Low B-J 13 0 13 0 SIS 7a Low Low None Low B-J 216 0 216 0 SIS 7a Low Low None Low C-F-I 9 0 7 0 CSS 7a Low Low None Low C-F-I 12 0 12 0 AFW 7a Low Low None Low C-F-2 0(4) 0(4) 130 0 Notes

1. Systems were described in Table 3.1-2 of the original submittal, with the exception of AFW - Auxiliary Feedwater. This ASME Code Class 2 system consists of 8 segments with 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. 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. 1 of the 3 welds selected for RI-ISI is the surge line elbow and is not counted as part of the weld count.
4. Due to a change in ASME Section XI Code criteria, 3" NPS Class 2 auxiliary feedwater piping was added to the ISI Program, and therefore the RI-ISI Program, for the first time during the third ISI interval. As such, there were no welds associated with this piping during the original RI-ISI application.

ATTACHMENT 2 PRA QUALITY REVIEW FOR RI ISI APPLICATION The Diablo Canyon PRA (DCPRA) is a living PRA, which is maintained through a periodic review and update process. The sections below discuss the review process that the DCPRA had under gone, the status of the disposition/resolution of the findings from the reviews and the impact of those outstanding/open issues from the reviews on the results and conclusions of this study.

Westinghouse Peer Review (Certification)

Peer Review (Certification) of the DCPP PRA model, using the WOG Peer Review Certification Guidelines, was performed in May 2000 and the final report for the peer review was published in August 2000 [Reference 11.

On the basis of its evaluation, the Certification Team determined that, with certain facts and observations (F&Os) addressed, the technical adequacy of all elements of the PRA would be sufficient to support risk significance evaluations with defense-in-depth input relative to the requested Emergency Diesel Generator completion time (CT) extension from the NRC during that time period.

The two "A" F&Os, related to the human reliability analysis (HRA) were addressed by upgrading the methodology used for the evaluation. The upgraded HRA analysis was recently subjected to a focused peer review.

A discussion of this focused HRA peer review is provided below.

The "B" F&Os from the WOG Peer Review were addressed during model updates in support of the EDG Completion Time Extension (CTE) license amendment request (LAR), the LAR effort to extend the Completion Times (CTs) for several emergency core cooling system (ECCS) components, and the MSPI calculations. The updated DCPRA model in which issues related to the "B" F&Os from the WOG Peer Review were addressed is the DC01 PRA model. There are no outstanding issues ("B" F&Os) from the WOG Peer Review.

DCPRA Gap Analysis In addition to the WOG Peer Review, three recent limited scope and independent assessments of the DCPP PRA Level 1 and Level 2 PRA models have been performed by leading industry PRA experts (i.e., Gap Analyses) to support several risk-informed applications, including the MSPI calculations and DCPP's transition to the National Fire Protection Association (NFPA) 805 Standard [Reference 2].

Self-Assessment of DCPRA Level 1 Internal Events A self-assessment of the Diablo Canyon Level 1 Internal Events PRA was performed by ERIN Engineering and Research, Inc. and the results were published in December 2006 (Reference 3) and then updated in January 2008 (Reference 4). The self-assessment was done with respect to the high level requirements (HLR) and supporting requirements (SR) in the ASME PRA Standard RA-Sb-2005, accounting for NRC interpretations of these requirements per Appendix A and Appendix B of Regulatory Guide 1.200 (Reference 5). One aim of the self-assessment is to identify SR for which the DCPP PRA may not meet the ASME PRA STD RA-Sb-2005 Capability Category II requirements. This category is generally viewed, for a given SR as sufficient capability for most currently envisioned risk-informed applications. The self-assessment did not include the determination of whether the DCPP PRA met the requirements for Large Early Release Frequency.

Table I summarizes the disposition/recommended action associated with the SR resulting from the self-assessment, and determines whether the issue associated the SR has any impact on the application. There are no open issues that would impact the results and conclusion of this evaluation.

Review and Re-evaluation of DCPP Internal Floodine Hazards The Diablo Canyon Internal Flood PRA (Reference 6) was reviewed by Scientech/Jacobsen Engineering (Reference 7) to identify any specific weaknesses in its approach or implementation which might impair its ability to be used for risk informed decision making. The approach for the review was to compare the method of implementation and documentation of the existing Internal Flooding PRA with the requirements of the ASME PRA standard Addendum B (March 17th 2005 Draft) (Reference 8).

Most of the review comments/finding was on the lack of documentation or related to the approaches not affecting the offsite power sources such as the 230kV power supply.

Table 2 summarizes the issues/deficiencies from the review and recommendations for improvement in some of the areas of the DCPP Internal Flooding PRA for the PRA to meet at least the Capability Category II requirements. The expected impact of these issues on the application is also provided. There are no open issues that would impact the results and conclusions of this evaluation.

Human Reliability Analysis Peer Review To address the findings and observations of an earlier peer review [in particular, the Human Reliability Analysis (HRA) portion] of the PRA (Reference 9), an upgrade of the HRA was performed (Reference 10). A follow-on peer review of the HRA was needed (required by ASME PRA Standard) and was performed by ABS Consulting, Inc. The findings were published in July 2007 (Reference 11). This peer review identified eighteen elements that did not meet category level II of the PRA Standard. Seven of these were documentation issues or did not affect the results of the application. The remaining eleven were subsequently resolved in an updated HRA Analysis (Reference 12). The updated HRA was incorporated into a sensitivity analysis to assess their impact on the RI-ISI application.

Level 2 Peer Review The level 2 peer review comments were reviewed for impact on the RI ISI results and conclusions. Since that Level 2 model was not used and a conservative estimate of LERF is used in the analysis, the Level 2 issues have no impact on the results and conclusions of this application.

Evaluation of Impacts of Open Issues and F&Os on RI ISI Analysis and Results The RI ISI evaluation relies specifically on the following PRA input and results to derive the risk rankings:

  • Initiating Event Frequencies o Large LOCA (LLOCA) o Medium LOCA (MLOCA) o Small LOCA Non-isolable (SLOCAN) o LOCA Outside Containment (VDI and VSI)

" Condition Core Damage Probability from the Following IE Sequences o LLOCA CCDP o MLOCA CCDP o SLOCAN CCDP o SLBI CCDP (Steam Line Break Inside Containment) o LPCC CCDP (Total Loss of CCW)

" Core Damage Frequency (CDF) With Mitigating Systems Failures o CDF with No RWST o CDF with No Charging Pumps o CDF with No AFW and No RWST o CDF No SI

" Component Failure Probability o Check Valve Fail to Close o MOV/AOV Fail to Close The open Internal Flood model issues do not significantly impact.the results since internal flood scenarios are not used in the RI ISI ranking application. The Level 2 open F&Os also do not impact the results

because the Level 2 model results are not used in the analysis, only conservative estimates of the LERF values are used. The open HRA F&Os have been resolved and HRA changes incorporated into a separate Riskman PRA model to evaluate the sensitivity of the RI-ISI application on HRA gaps. This sensitivity model (entitled DCO1ISI - Reference 13) also incorporates updated component failure and initiating event frequency data.

The sensitivity model resulted in the following consequence case rank changes:

"M M...

I111 .llll I Ilkl*..*l I! I I .II *"It" *,nIIIIl'll I*

Case Rank CCDP CLERP Rank CCDP CLERP Segments separated from the RCS by one ACC-4 M 4.50E-05 4.50E-07 H 1.48E-04 1.48E-06 intact check valve and break is inside containment.

ACC-7 M L All segments that, if failed, would disable SI 7.40E-06 7.40E-08 8.51E-07 8.51 E-09 injection paths to the RCS.

CS-1 M 4.50E-05 4.50E-07 H 1.48E-04 1.48E-06 CS Pump 1-1 Discharge header, from the pump to the normally closed MOV 900 1A CS-2 M 4.50E-05 4.50E-07 H 1.48E-04 1.48E-06 CS Pump 1-2 Discharge header, from the pump to the normally closed MOV 9001B Charging pumps suction header - from normally closed MOVs 8805A and B and check valve 8924/

CVCS-5 M 6.03E-06 6.03E-08 L 3.26E-07 3.26E-09 Charging pumps suction header - from check valve 8924, to reducer after valve 8925, normally closed MOV-8804A (RHR supply line), reducer between S2-1356 and S2 41-4, including the suction lines to the pumps As a result of incorporating the updated HRA and PRA model data, three consequence case rankings increase and two decrease relative to the DCO1 production model. The overall impact on the RMSI application is limited and the PRA model changes that drive the rank differences will be incorporated into the production model in time for the next scheduled DCPP RI-ISI request (ref. DCPP SAPN 50451698).

The current open issues and F&Os were reviewed for impact against the above items and results summarized in the table below.

SUMMARY

OF PRA MODEL INPUT AND RESULTS IMPACT ON RI IS[ APPLICATION Sensitivity Impact on RI of RI ISI ISI Results PRA Results to Significant Open and RI ISI Model PRA Input and Results Value PRA Input Issues and F&Os Conclusions Discussion Large LOCA (LLOCA) IE 2.43E-06 High None None IE data has been updated in sensitivity analysis.

Limited -

Increases rank of CS-1, CS-2 and ACC-4 Medium LOCA (MLOCA) IE 1.46E-04 High None segments IE data has been updated in sensitivity analysis.

Small LOCA Non-isolable (SLOCAN) IE 1.27E-03 Low None None IE data has been updated in sensitivity analysis.

LOCA Outside Containment (VDI and VSI) lEs 1.00E-05 High None None IE data has been updated in sensitivity analysis.

HEPs in model have been updated for sensitivity analysis. Internal Flood model has insignificant LLOCA CCDP 2.88E-02 Medium HR and IF Issues Insignificant impact on results.

HEPs in model have been updated for sensitivity analysis. Internal Flood model has insignificant MLOCA CCDP 2.95E-02 Medium HR and IF Issues Insignificant impact on results.

HEPs in model have been updated for sensitivity analysis. Internal Flood model has insignificant SLOCAN CCDP 2.06E-04 Medium HR and IF Issues Insignificant impact on results.

HEPs in model have been updated for sensitivity SLBI CCDP (Steam Line Break analysis. Internal Flood model has insignificant Inside Containment) 9.06E-06 Low HR and IF Issues Insignificant impact on results.

HEPs in model have been updated for sensitivity analysis. Internal Flood model has insignificant LPCC CCDP (Total Loss of CCW) 1.35E-03 Low HR and IF Issues Insignificant impact on results.

HEPs in model have been updated for sensitivity analysis. Internal Flood model has insignificant CDF with No RWST 4.27E-03 Medium HR and IF Issues Insignificant impact on results.

HEPs in model have been updated for sensitivity analysis. Internal Flood model has insignificant CDF with No Charging Pumps 8.23E-05 Low HR and IF Issues Insignificant impact on results.

HEPs in model have been updated for sensitivity analysis. Internal Flood model has insignificant CDF with No AFW and No RWST 4.93E-01 Low HR and IF Issues Insignificant impact on results.

HEPs in model have been updated for sensitivity analysis. Internal Flood model has insignificant CDF No SI 7.14E-05 Low HR and IF Issues Insignificant impact on results.

Check Valve Fail to Close 5.38E-05 Medium None None Data has been updated in sensitivity analysis.

MOV/AOV Fail to Close 2.43E-03 Low None None Data has been updated in sensitivity analysis.

References

1. Diablo Canyon Power Plant Probabilistic Risk Assessment Peer Review Report, Final Report, August 2000
2. National Fire Protection Association (NFPA) 805 Standard
3. Diablo Canyon Power Plant PRA Self-Assessment (Draft Report), ERIN Engineering and research, December 2006.
4. Diablo Canyon Power Plant PRA Self-Assessment, ERIN P0114060001-2717 Ri, January 2008.
5. Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk- Informed Activities," U.S. NRC dated, January 2007.
6. "PRA Internal Flooding Analysis," Calculation File F.4, Revision 1
7. "Review and Reevaluation of Specific Issues of internal Floods Analysis"
8. American Society of Mechanical Engineers (ASME) RA-Sb-2003, Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated March 17, 2005 (draft)
9. Davis, Ernest G., "Human Action Analysis - Failure Likelihood and Range Factor Calculation",

Calculation file number: G.2, revision 5, 392 pages, undated but after May, 2006

10. Diablo Canyon Follow-On Peer Review of HRA Update, Final Report, R-1736044-1728, July 31, 2007.
11. Diablo Canyon Follow-On Peer Review of HRA Update, Final Report, R-1736044-1728, July 31, 2007.
12. Calculation File G.2, Revision 6, "Failure Likelihood and Range Factor Calcs," Mar 2010.
13. PRA Calculation File PRAOO-05 Revision 2.
14. Risk Impact Analysis for DCPP 12P3, Attachments 4 and 5 (Inservice Engineering, March, 2010)

Table 1. Summary of Suggested Disposition Actions from the DCPRA Gap Analysis (See Table 1 of Reference 7)

Applicable Description and Suggested Disposition Action Expected Impact on Application ASME SRs IE-A7 IE-A7 is met at Capability Category I; precursors are not directly Calc File H. 1.6 updated to include factored into the model. However, this may be a pessimistic discussion of screening of precursor assessment, since insights gained from past precursors has been events. Documentation Issue and no incorporated, so Capability Category II could be appropriate. The set impact is expected of initiating events modeled is believed to adequately represent the spectrum of applicable industry experience, and it is unlikely that not meeting Capability Category II for this SR would have an impact on applications of the PRA. Consider adding a discussion of how initiating event precursors should be addressed to either the H. 1.6 calc or to PRA update guidance.

IE-A10, IE- IE-AIO is Not Met. The treatment of dual unit initiators should be ASW for Unit 2 only credited if U2 B5, SC-A4a, reviewed, and the documentation of the basis for the current EDGs are operable and can support SY-A1I1 treatment, or an update, should be developed, pumps. Vital power cross tie not currently credited. Calc File H. 1.6 updated to include discussion of plant response to dual-unit initiators.

Documentation Issue and no impact is expected SC-A6, SC- While SR SC-A6, SC-B 1, SC-B3 are judged to be met, the issues in Both ATWS issue (F&O DA-7) and BI, SC-B3 C-significance F&Os DA-7 and TH-4 might have significance to PTS issue (under TH-4) resolved.

particular applications. The impact of these should be considered on Calc File E. II updated to reflect an application-specific basis until resolved changes. No impact on application.

SY-A20 To meet SR SY-A20, a confirmation that credited SSCs are able to Equipment qualification discussion operate in all modeled accident scenarios, including those where to be addressed in Calc File E. 17.

SSC design basis conditions may be exceeded, is needed. No impact on application.

HR-D4 HR-D4 is met with one exception, lack of an established HRA was updated and this maximum credit for recovery in the pre-initiator HEPs. finding closed.

Although a maximum credit is not assigned, excessive credit is not taken for recovery. Therefore, this SR has been judged to be adequately met. However, this issue could easily be addressed in the documentation.

HR-G4 HR-G4 does not appear to be met. The bases for HEP timing success HRA was updated and this criteria analyses are not adequately specified in Calc G.2; times are finding closed..

specified but the bases for the times are unclear in the calc. (They may be documented in the HRA Calculator). [This assessment is based on information available prior to the re-peer review of the HRA.]

HR-G5 HR-G5 does not appear to be met. The validation of human action HRA was updated and this timing is unclear. Calc G.2 refers to operator interviews for required finding closed.

times, but it is unclear as to what this covers. [This assessment is based on information available prior to the re-peer review of the HRA.]

DA-D2 DA-D2 is currently NA since there are no instances of failure events SR N/A. No impact with no applicable generic data. Consideration should be given to developing a process for estimating data for which there is no generic data source, consistent with the DA-D2 requirements, for future application.

Table 1. Summary of Suggested Disposition Actions from the DCPRA Gap Analysis (See Table 1 of Reference 7)

Applicable Description and Suggested Disposition Action Expected Impact on Application ASME SRs DA-D7 DA-D7 is currently NA since there are no instances where existing SR N/A. No impact plant experience data are no longer applicable. Consideration should be given to developing a process/guidance for dealing with data that are no longer applicable, consistent with the DA-D7 requirements, for future application.

QU-D4 QU-D4 is Not Met. Consideration should be given to adopting a Discussion of the review of non-sampling process for review of non-dominant sequences as part of significant sequences has been the model quantification. included in Calc File C.9.

Documentation issue and no impact on application.

TABLE 2. Summary of DCPP Internal Flood Analysis Areas for Improvement (Table 2-1 of Reference 4)

Supporting Summary of Existing Recommendation for Expected Impact Requirement DCPP Internal Flood PRA Improvement to meet on Application from ASME Approach (Reference 3) Standard Std and Nature of Associated Deficiency with Respect to ASME Std IF-A4 A plant walk down was conducted Analysis needs to be brought up to Documentation issue IF-B3a as part of the Rev 0 analysis to date by repeating procedure and no impact on IF-C9 collect additional information to performed for Rev 1 analysis. application is expected IF-A3 confirm previous documentation Since that last comprehensive and judgments on the flood walkdown was performed in 1991 (and sources and potential impact. This not well documented) it is was documented by photographs recommended that it is repeated and of important equipment. For Rev documented using a set flood area 1 analysis an additional walkdown walk down sheets and checklists.

was conducted (Attachment 6 -

Reference 3) to confirm information used in the intake structure analysis.

IF-B1 Major flood sources in each area See recommendation for creating Documentation issue are identified by type (e.g. water flood area information sheets in IF-A4 and no impact on piping high/ moderate energy, which identified requisite information, application is expected steam piping high energy) in table F.4-2. However specific systems, pipe sizes or external flood sources are not identified.

Potential in leakage is not explicitly identified although it can be inferred from the propagation paths "to" column (Note in some cases the potential for propagation is identified without describing the specific route (doorway opening etc. Where pathway is described no explicit reference is given (e.g. door number)

IF-Bib Table F.4-2 (Reference 3) Recommend defining a set of Flooding events are IF-C5 includes a screening process. qualitative and quantitative screening not significant IF-C5a However the general screening criteria consistent with the ASME contributors to CDF.

criteria used are not well defined standard and indicating for each Impact on conclusions and justified and in some cases specific flood area which particular of the current include judgmental credit for criteria is applicable, application will be isolation of sources before negligible sirfce offsite damage/ propagation can occur and onsite power and /or drainage capacity. The sources not affected.

containment is screened out on the basis that it is designed for LOCA and high energy line breaks in containment (section F.4.3).

IF-Cl Table F.4-2 identifies the flood In flood area information sheets (see Only areas where IF-C3b propagation paths from the source IF-A4) document potential propagation significant area to an adjacent area ( but no paths thru cable penetrations as well accumulation leading further) Section F.4.3.2 provides a as doors and HVAC ducts to structural failure of general discussion Section F.4.3.2 barrier elements are provides a good general Review flood analysis to identify cases the SI Pump room and description for each building where flood accumulation may occur Charging Pump Room.

(turbine, intake, auxiliary and fuel (or has not been ruled out) and In these cases one can handling) of the flood propagation determine if consequences of barrier assume that barrier pathways to their ultimate point of element challenges (e.g. doors or failures may lead to accumulation penetration seals) may result in a damage in adjacent plant impact which has not been areas where Did not see any reference to addressed in the current flood appropriate.

analysis of structural failures in analysis. Ifso perform engineering the analysis although this analysis to determine if the barrier Flooding events are

TABLE 2. Summary of DCPP Internal Flood Analysis Areas for Improvement (Table 2-1 of Reference 4)

Supporting Summary of Existing Recommendation for Expected Impact Requirement DCPP Internal Flood PRA Improvement to meet on Application from ASME Approach (Reference 3) Standard Std and Nature of Associated Deficiency with Respect to ASME Std probably because the potential for element will withstand the loading not significant significant flood accumulation in contributors to CDF.

most cases is minimal given the Performing the above will satisfy the Impact on conclusions plant design Cat II requirements. of the current application will be The only evidence of random In order to satisfy Cat III requirements negligible since offsite barrier element failures being the random failure of any barrier and onsite power considered is in respect of the elements identified as being sources not affected.

ASW room drain check valves, challenged and with significant consequences of failure will need to be addressed. ( See Westinghouse F&O type C ) For area which may be susceptible to high energy line breaks determine if barriers and barrier elements will be challenged by over pressure and determine consequences of failure.

IF-C2a Automatic or operator responses Where such features form part of the Consequences of non IF-C6 to terminate floods are argument for screening or evaluating isolation in a timely summarized in the discussion of flood scenarios this information should manner were found to flood location and scenario be provided, be potentially evaluations provided in section Specifically recommend evaluating the significant for SCW, F.4.3.2 Table F4 Scenarios 45, reliability of actions credited in the CCW, RWST supply 46, 53, 54 appears to credit scenarios 45, 46, 53, 54, 69, 83 and floods and AFW pump manual action for isolation in order 84. room floods in terms of to screen (although it is not clear controlling the potential whether the consequences of non for flood propagation or isolation are significant). No gross system impact. A discussion of flood indication, screening analysis is timing or means of isolation is proposed which does provided, not credit isolation, except in the case of the SCW where the flooding rate relative to the volume required to cause damage was judged to be such that the time available would be many hours.

Flooding events are not significant contributors to CDF.

Impact on conclusions of the current application will be negligible since offsite and onsite power sources not affected.

IF-C3 Equipment susceptibility to Need to include potential damage to Impact of various types of flood hazard are junction boxes treatment due to spray high/moderate energy identified in table F.2-12 of the and submergence. line breaks original DCPP flood PRA. In (HELB/MELB) need to summary this table indicates that be considered for all electrical components except Capability Category I1.

cables are assumed to be An update of the susceptible to flood accumulation Internal Flooding and spray. High energy jet Analyses should

TABLE 2. Summary of DCPP Internal Flood Analysis Areas for Improvement (Table 2-1 of Reference 4)

Supporting Summary of Existing Recommendation for Expected Impact Requirement DCPP Internal Flood PRA Improvement to meet on Application from ASME Approach (Reference 3) Standard Std and Nature of Associated Deficiency with Respect to ASME Std impingement may cause damage evaluate and document to all electrical components the impact of including cables. No reference to HELB/MELB.

junction box qualification/damage is given and the treatment needs Impact on conclusions to be checked. It is not clear that of the current high energy line break effects application will be have been considered in the negligible since offsite Revision 1 update and onsite power sources not affected.

IF-C3c The results of engineering Need to identify location of flooding An update of the calculations of maximum flood calculations relied upon in the analysis Internal Flooding heights reported in DCM T-20 and review underlying basis to ensure Analysis should (see table F.4.4) are used in the consistency with PRA requirements ( determine the study. For example a maximum e.g. no restrictions on maximum crack applicability of design flood height of 3" is cited as the size or assumptions about isolation calculations cited in the reason for lack of flood within specific time) existing Internal propagation from the AFW TDP Flooding study.

pump room to the AFW MDP pump rooms. When this Impact on conclusions reference was reviewed the of the current calculation referred to was not application will be apparent negligible since offsite and onsite power sources not affected.

IF-C4 Flood scenario development is Screening analyses proposed which No expected to impact generally accomplished in section does address impact of isolation of the conclusions of the F4.5.1. However it is not clear the flood source. Further detailed current application analysis has recognized the analyses may be needed if this since offsite and onsite consequences of flood isolation conservative analysis shows high risk power sources not on system availability That is contribution, affected.

isolation of an AFW system flood may require the CST source to all pumps to be isolated depending upon (the break location).

Isolation of a CCW flood may require partial isolation of the CCW system IF-C5a It appears that DCPP analysis Further examination of the reliability of An update of the (Table F.4-2 item 23) credits the isolation system, the timing Internal Flooding isolation of a large turbine building available for operator action, the Analysis will re-flood prior to propagation to the integrity and reliability of the doors and examine the Turbine DG corridor or the fuel oil pump drain check valves which protect the Building flood room vaults via drains, and the EDG rooms, the fuel oil pump vaults scenario(s). Since it is 12kV room due to the automatic and the 12kv SWGR room as well as extremely unlikely that condenser mitigating features. any drainage paths to the outside, is the EDG will be This qualitative argument is used warranted in order to screen this affected by TB flooding to screen out all propagation scenario (Although extremely unlikely events, this issue scenarios from the turbine this scenario could lead to a loss of would have building. the EDGs and loss of offsite power). insignificant impact on the conclusion of the current application.

IF-D3 Flood scenanos are grouped as If the consequences of isolation of the This issue has been follows: CCW or AFW is potentially more partially resolved. Not FL1 - All CCW floods significant than currently identified expected to impact the FL2 - Charging suction header (See IF-E5a) flood scenarios may conclusions of the

TABLE 2. Summary of DCPP Internal Flood Analysis Areas for Improvement (Table 2-1 of Reference 4)

Supporting Summary of Existing Recommendation for Expected Impact Requirement DCPP Internal Flood PRA Improvement to meet on Application from ASME Approach (Reference 3) Standard Std and Nature of Associated Deficiency with Respect to ASME Std floods need to be broken up in order to current application FL3 - AFW OR Fire Water Floods recognize specific consequences since offsite and onsite in AFW MDP Room associated with different break power sources not locations in order to meet cat II or III affected.

requirements.

IF-E5a Only three non screened flood Need to address potential impact of Although a screening scenarios were developed for flooding on HEPs included in the HEP value was used in quantification of CDF. Only the internal events PRA including an the scenario, this is not CCW scenario which credited assessment of degradation of expected to impact the isolation (prior to system instrumentation and access for local conclusions of the depressurization ) and in this actions current application analysis a 10% probability of Need to perform more robust since offsite and onsite failure to isolate was assumed justification of flood isolation power sources not without any justification on the probability used for CCW floods and affected.

basis of flood indication, event impact.

timing, and means of isolation. In addition the analysis does not appear to address the consequences of conducting isolation which presumably would be lead to at least partial loss of the system. None of the three flood scenario analyses appear to have addressed the potential degradation on operator errors modeled in the PRA associated with the flooding event. (see IF-D3)

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