RC-17-0123, Relief Request RR-4-13, Use of a Risk-Informed Process as an Alternative for the Selection of Class 1 and Class 2 Piping Welds

Relief Request RR-4-13, Use of a Risk-Informed Process as an Alternative for the Selection of Class 1 and Class 2 Piping Welds

ML17303B183
Person / Time
Site:	Summer
Issue date:	10/30/2017
From:	Lippard G South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RC-17-0123, RR-4-13
Download: ML17303B183 (28)

Text

George A. Lippard Vice President, Nuclear Operations A SCAN A COMPANY October 30, 2017 RC-17-0123 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 RELIEF REQUEST RR-4-13, USE OF A RISK-INFORMED PROCESS AS AN ALTERNATIVE FOR THE SELECTION OF CLASS 1 AND CLASS 2 PIPING WELDS In accordance with the provisions of 10 CFR 50.55a(z)(1), South Carolina Electric & Gas Company (SCE&G), acting for itself and as an agent for South Carolina Public Service Authority (Santee Cooper) hereby submits the attached request for using an alternative to the inservice testing requirements of the ASME code. SCE&G has determined that the proposed alternative would provide an acceptable level of quality and safety.

A detailed description of the proposed alternative, including basis for relief, is enclosed with this letter SCE&G requests NRC review and approval of this request by August 15, 2018 in order to support planning for refueling outage (RF-24) which is schedule to start in the Fall of 2.018.

SCE&G is submitting the attached relief request in accordance with 10CFR50.55a(z)(1).

V. C. Summer Nuclear Station

• P. 0. Box 88

• Jenkinsville, South Carolina

• 29065

• F (803) 941-9776

• www.sceg.com

Document Control Desk CR-16-01194 RC-17-0123 Page 2 of 2 Should you have any questions, please call Bruce L. Thompson at 803-931-5042.

BJD/GAL/wk

Enclosure:

RELIEF REQUEST RR-4-13 : Summary Statement of VCSNS Unit 1 PRA Model Capability for use in Risk-Informed Inservice Inspection Program Licensing Actions. : VCSNS - Inspection Location Selection Comparison Between Previously Approved and Revised RI-ISI Program by Risk Category Very truly yours, George A. Lippard cc/w:

K. B. Marsh S. A. Byrne J. B. Archie N. S. Cams G.J. Lindamood W. M. Cherry C. Haney S. A. Williams NRC Resident Inspector K. M. Sutton NSRC RTS (CR-16-01194)

File (810.19-2)

PRSF (RC-17-0123)

J. H. Hamilton

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 1 of 10 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ENCLOSURE Relief Request RR-4-13 1

Subject VCSNS Unit 1 is proposing an alternative to the requirements of the inspection and examination requirements of Class 1 and 2 piping welds specified by the ASME Code, Section Xl, Tables IWB-2500-1 and IWC-2500-1. 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 Ten-Year ISI Interval.

2 ASME Code Component(s) Affected All Code Class 1 and 2 piping welds previously subject to the requirements of ASME Section XI, Table IWB-2500-1, Examination Categories B-F1 and B-J, and Table IWC-2500-1, Examination Categories C-F-1 and C-F-2.

3

Applicable Code Edition and Addenda

The applicable Code Edition and Addenda for the Fourth Ten-Year Inservice Inspection (ISI) Interval at V.C. Summer Nuclear Station (VCSNS) is the 2007 Edition with 2008 Addenda of ASME Section XI.

The station is in its fourth 10 year interval effective from January 1, 2014, through and including December 31, 2023.

4

Applicable Code Requirement

The selection process for Code Class 1 and Code Class 2 pipe welds to be examined in the Fourth Ten-Year ISI Interval is prescriptively determined in accordance with ASME Section XI Table IWB-2500-1, Examination Categories B-F1 and B-J, and Table IWC-2500-1, Examination Categories C-F-1 and C-F-2.

1 Note that although Examination Category B-F welds are included in the RI-ISI program for the consideration of all possible degradation mechanisms, Alloy 600/82/182 examinations in the Third Interval were conducted per Code Cases N-722-1 and N-770-1. In the Fourth Interval, these examinations will be performed in accordance with the versions of the applicable Code Cases that are referenced in the published version of 10CFR50.55a.

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 2 of 10 5

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 Ten-Year ISI Interval.

Use of the risk-informed selection process has been shown to reduce the probable frequency of core damage and large early release when compared to the prescriptive deterministic selection method. The methodology will continue to be based on Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Rev. B-A, with identified differences and additional guidance taken from ASME Code Case N-578-1 and ASME Section XI Nonmandatory Appendix R.

6 Proposed Alternative and Basis for Use As an alternative to the Code Requirement, a Risk-Informed process will be used for selection of Class 1 and Class 2 piping welds for examination.

=

Background===

In 2002, a risk-informed (RI) methodology for the inservice inspection of Class 1 and 2 piping welds was applied at the V.C. Summer Nuclear Station based on Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Rev. B-A, with identified differences and additional guidance taken from ASME Code Case N-578. The original RI-ISI template, "Risk-Informed Inservice Inspection Program Plan, V.C. Summer Nuclear Station, Rev. 0,"

was submitted to the NRC for approval as Attachment 2, Relief Request RR-II-07, to letter RC-02-0161, dated September 16, 2002, and supplemented in a letter to the NRC dated January 29, 2003. Based upon the information provided in the RI-ISI template and supplemental submittal described above, the request to implement the RI-ISI methodology on Class 1 and 2 piping welds was approved by the NRC for VCSNS's Second Ten-Year ISI Interval in an NRC SER dated May 12, 2003.

In 2004, the RI-ISI application was evaluated and updated in conjunction with the Third Interval ISI Program Update. This resulted in the generation of Relief Request No. RR-III-02 which addressed continued use of the RI-ISI application during the Third Interval. By letter RC-04-0148 to the NRC on September 8, 2004, South Carolina Electric and Gas submitted Relief Request RR-III-02 for VCSNS requesting relief from the ASME Section XI Code examination requirements of Class 1 and 2 piping weld (Examination Categories B-F, B-J, C-F-1 and C-F-2) inservice inspections by continuing implementation of their RI-ISI Program. Relief Request RR-III-02 was approved by the NRC in a Safety Evaluation Report dated September 6, 2005 (ML052300616).

For the Fourth ISI Interval, which commenced on January 1, 2014, VCSNS intended to resubmit the RI-ISI Program. However, since VCSNS did not submit its RI-ISI Program before or during the first period, conventional Section XI rules were applied to Examination Categories B-F, B-J, C-F-1 and C-F-2 during this period. The first period of the Fourth ISI Interval was completed on June 1, 2017. Upon NRC approval of the RI-ISI Program for the Fourth Interval, VCSNS will prorate examinations in these examination categories accordingly for the remainder of the interval.

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 3 of 10 RI-ISI Living Program Evaluations and Updates The original VCSNS RI-ISI Program submittal to the NRC contained the following statement related to the evaluation/update process:

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, piping segments will be reviewed and Risk Ranking adjusted as necessary on an ASME period basis. In addition, significant RI-ISI changes may require more frequent adjustment as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant specific feedback.

The requirement to perform living program evaluations and updates of the RI-ISI Program on a period basis is still applicable. Guidelines for the performance of these living program evaluations and updates are provided in NEI 04-05, Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems, published April, 2004. In accordance with NEI 04-05, the following aspects were considered during the periodic reviews for VCSNS:

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 The updated RI-ISI Program resulting from these periodic evaluations is the subject of this proposed alternative.

During the review after the First Period of the Third Interval, the following changes were identified and incorporated into the RI-ISI Program:

1. The Consequence Evaluation was updated to reflect the latest revision of the PRA.

2. The Risk Ranking Summary, Matrix, and Report were updated to incorporate the change of Risk Categories for 90 segments due to the PRA model update as identified in the RI-ISI Evaluation.

3. The increase in Risk Ranking for 32 Main Steam segments resulted in 143 welds increasing from Low Risk to Medium Risk. A Medium Risk Category requires the

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 4 of 10 selection of 10% of the population. Therefore, fifteen additional welds were selected for examination.

4. The Risk Impact Analysis was updated to reflect the new Upper Bound Conditional Core Damage Probability (CCDP) and Conditional Large Early Release Probability (CLERP), the revised Risk Categories, and the additional Main Steam elements selected for examination. Risk Impact remained negative with Probability of Detection (POD) considered, and negligible without POD.

5. The Risk Ranking and Risk Impact Analysis were updated to reflect the treatment of Primary Water Stress Corrosion Cracking (PWSCC) in a separate program, similar to the treatment of Flow Accelerated Corrosion (FAC). As a result of this update, there was the potential to reduce the elements selected for examination by 4. However, VCSNS decided that the selections would remain the same until implementation of the SAP-1281 PWSCC Program has been completed and implemented.

As a result of these changes, the number of elements selected for inspection increased from 94 to 109. The increase was due to the change in Consequence Category to 32 Main Steam segments. The potential decrease in examinations due to the treatment of PWSCC in an independent program was deferred until the effectiveness of the independent program could be verified. The total risk (both Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)) continued to be lower than that under the original deterministic Section XI program when POD is considered, and essentially unchanged without consideration of POD.

During the review after the Second Period of the Third Interval, the following change was identified:

1. During the Second Period the PRA model changed from Version 5 to Version 6c.

As a result of this change, the Conditional Core Damage Probability (CCDP) of the 32 Main Steam segments that changed as a result of the First Period update decreased sufficiently to change those segments from Risk Category 4 (Medium Risk) back to Risk Category 6a (Low Risk). This resulted in a potential return of the number of Main Steam examinations to what was previously required at the start of the Third Interval. However, VCSNS decided to conservatively keep the RI-ISI examinations as they were until the end of the Third Interval.

During the review after the Third Period of the Third Interval, the following changes were identified and incorporated into the RI-ISI Program:

1. Consideration was given to the selection of welds 1-4100A-26BC and 1-4200A-22BC for RI-ISI examination. This recommendation was the result of Condition Report CR-13-02110 which identified periodic thermal transients in the RCS Cold Leg Safety Injection piping. This piping had already been identified as being subject to thermal transients in the RI-ISI Degradation Mechanism Evaluation and welds 1-4106A-8, 1-4106A-9, 1-4202A-16 and 1-4202A-17 have been selected for examination. Therefore, the criteria of the RI-ISI methodology have been met. In

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 5 of 10 addition, welds 1-4100A-26BC and 1-4200A-22BC are small bore branch connection welds that are not conducive to examine by ultrasonic examination.

However, in order to address Condition Report CR-13-02110, VCSNS opted to schedule owner-elected augmented surface examinations on these welds outside of the RI-ISI Program.

2. During the Third Period the PRA model changed from Version 6c to Version 7. As a result of this change, The Consequence Rank for System/Train Loss identifiers SI B and SI C increased from Low to Medium. As a result, thirty-nine Chemical and Volume Control System (CVCS) segments changed from Risk Category 7a to Risk Category 6a. Since both Risk Category 6 and Risk Category 7 are Risk Rank Low, there is no change in Risk Rank and no additional examinations were required. However, the Consequence Evaluation and Risk Ranking Summary, Matrix and Report were updated to reflect the changes.

3. Per Version 7 of the PRA Model, the Upper Bound CCDP used in the Risk Impact Analysis is 6.25E-05 and the Upper Bound CLERP is 3.00E-08. The Risk Impact Analysis was updated to reflect the new Upper Bound CCDP and CLERP. The Risk Impact remained negative with POD considered, and negligible without POD.

4. During the Third Interval an assessment was performed comparing the ISI drawings to the ISI database and any resulting differences were reconciled. During the Third Period, the Risk Ranking Summary, Matrix and Report and Risk Impact Analysis were updated to reflect the reconciled ISI database information.

5. The Risk Ranking, Element Selection, and Risk Impact Analysis were updated to reflect the application of Code Case N-770-1. The examination of welds due to PWSCC is considered administratively during the RI-ISI application, but the Code Case N-770-1 program takes precedence. Therefore, welds subject to PWSCC are selected for examination per Code Case N-770-1 and examined under that program. Welds for which no other degradation mechanism has been postulated will be examined solely under the Code Case N-770-1 Program and were removed from consideration during the RI-ISI element selection process. Welds for which another degradation mechanism other than PWSCC has been postulated were considered for further examination in the RI-ISI application in the same population as those subject to the additional degradation mechanism. However, the Code Case N-770-1 augmented examination program is not changed by the RI-ISI application and will remain in effect.

6. The Risk Ranking and Risk Impact Analysis were updated to include PWSCC as a potential degradation mechanism for the following welds:

1-4100A-16(DM) 1-4100A-32(DM) 1-4200A-16(DM) 1-4200A-29(DM) 1-4300A-16(DM)

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 6 of 10 1-4300A-30(DM)

It was also noted that the following steam generator inlet and outlet dissimilar metal welds were inlayed with Alloy 690 material during steam generator replacement in 1994:

1-4100A-31 (DM)

S/G 2A Inlet 1-4100A-32 (DM)

S/G 2A Outlet 1-4200A-28 (DM)

S/G 2B Inlet 1-4200A-29 (DM)

S/G 2B Outlet 1-4300A-29 (DM)

S/G 2C Inlet 1-4300A-30 (DM)

S/G 2C Outlet Although these welds were administratively noted as potentially having PWSCC as a degradation mechanism in the RI-ISI Program, the Code Case N-770-1 Program superseded the RI-ISI Program for these welds during the Third Interval. During the Fourth Interval the susceptibility of these welds will be determined and their corresponding examinations will be performed in accordance with the version of Code Case N-770 that is referenced in the published version of 10CFR50.55a as discussed under Augmented Examination Requirements.

7. The Risk Ranking Summary, Matrix and Report and Risk Impact Analysis were updated to remove welds beneath weld overlays which are subject to a separate examination program.

As a result of incorporating these changes from the Third Period as well as the change identified in the Second Period, the number of elements selected for inspection decreased from 109 to 90.

The total risk (both CDF and LERF) remained decreased when compared to that under the original deterministic Section XI program when POD is considered, and essentially unchanged without consideration of POD.

During the review after the First Period of the Fourth Interval, the following changes were identified and incorporated into the RI-ISI Program:

1. In refueling outage RF22, weld 2-2104-6 had limited examination coverage. Weld 2-2204-6 was chosen to supplement the examination of weld 2-2104-6. This additional selection is documented in the Risk Ranking Summary and Report.

2. During the First Period of the Fourth Interval the PRA model changed from Version 7 to Version 8a. As a result of the change in PRA model, the Consequence Ranking changed for 58 Consequence IDs and the associated 82 segments. The Consequence Evaluation, Element Selections, Risk Ranking Summary, Matrix and Report and Risk Impact Analysis were updated to reflect these changes.

3. Per Version 8a of the PRA Model, the Upper Bound CCDP used in the Risk Impact Analysis is 2.63E-03 and the Upper Bound CLERP is 1.10E-03. The Risk Impact Analysis was updated to reflect the new Upper Bound CCDP and CLERP. The Risk Impact remained negative with POD considered, and negligible without POD.

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 7 of 10

4. During the RI-ISI Evaluation and Update for the First Period of the Fourth Interval it was determined that ASME Section XI, Nonmandatory Appendix R and Code Case N-578-1 should be referenced for guidance of the RI-ISI Program application.

As a result of incorporating these changes from the First Period of the Fourth Interval, the number of elements selected for inspection increased from 90 to 97. The total risk (both CDF and LERF) continued to be lower than that under the original deterministic Section XI program when POD is considered, and essentially unchanged without consideration of POD.

Risk Impact Analysis All issues identified in the Periodic Reviews have been incorporated into the Risk Ranking, Summary, Matrix and Report. 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 core damage frequency (CDF) and large early release frequency (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 when POD is considered.

The revised program represents an overall reduction of plant risk of -9.83E-09 in regards to CDF and -4.08E-09 in regards to LERF.

As indicated in the VCSNS Risk Impacts Results 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 or not the enhanced POD is credited for the RI-ISI examinations.

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 8 of 10 VCSNS Risk Impact Results System RiskCDF RiskLERF w/ POD w/o POD w/ POD w/o POD CS

-3.40E-09

-1.92E-09

-1.42E-09

-7.99E-10 EF

-2.40E-11 0.00E-00

-2.40E-11 0.00E-00 FW

-1.20E-11 0.00E-00

-1.20E-11 0.00E-00 MS

-5.26E-11

-5.26E-11

-2.20E-11

-2.20E-11 RC

-4.72E-09 3.29E-10

-1.97E-09 1.38E-10 RHR

-7.28E-10

-8.89E-11

-2.99E-10

-3.40E-11 SI

-8.80E-10 2.28E-09

-3.56E-10 9.65E-10 SP

-1.32E-11

-1.32E-11

-5.50E-12

-5.50E-12 SW Negligible Negligible Negligible Negligible Total

-9.83E-09 5.42E-10

-4.08E-09 2.43E-10 Augmented Examination Programs The following augmented inspection programs were considered during the RI-ISI application:

The augmented examination program for flow accelerated corrosion (FAC) per Generic Letter 89-08 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 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 inspection program for the service water intake and piping is addressed in Procedure ES-505, "Service Water System Corrosion Monitoring and Control Program."

This procedure is relied upon to manage this damage mechanism (i.e., Microbiologically Influenced Corrosion (MIC) and Pitting (PIT)) but it 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 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 were performed during the Third Interval in accordance with Code Case N-770-1 which was relied upon to manage the damage mechanism of PWSCC but was not otherwise affected or changed by the RI-ISI Program. Note that welds selected for examination in

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 9 of 10 accordance with Code Case N-770-1 were considered as part of the RI-ISI population such that they were evaluated for other potential degradation mechanisms. However, they were 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 10CFR50.55a. Per the Final Rule for 10CFR50.55a dated August 17, 2017, Code Case N-770-2 is the current applicable version.

Additional Examinations Whenever RI-ISI examinations reveal flaws or relevant conditions exceeding acceptance standards, additional examinations shall be performed during the current outage using the criteria of ASME Section XI, Nonmandatory Appendix R, Section R-2430.

Proposed ISI Program Plan Change Request VCSNS requests to submit RI-ISI application in accordance with 10CFR50.55a(z)(1). A comparison between the proposed RI-ISI program and the previously approved RI-ISI Program is provided in Attachment 2.

VCSNS completed the First Period of its Fourth Interval on June 1, 2017. Up until this point, 29.1% of Examination Category B-F, B-J, C-F-1, and C-F-2 weld examinations have been completed in the interval. Beginning in the Second Period of the Fourth Interval, the examinations determined by the RI-ISI process will replace those selected per ASME Section XI criteria. Since 29.1% of the examinations have been completed thus far in the Fourth Interval, 70.9% of the RI-ISI examinations will be performed during the remaining refueling outages in the Fourth Interval. Note that this is the same proration approach that was requested and approved by the NRC in Relief Request RR-II-07. Subsequent ISI intervals will implement 100% of the examination locations selected per the RI-ISI program.

Examinations shall be performed during the interval such that the period examination percentage requirements of ASME Section XI, paragraphs IWB-2412 and IWC-2412 are met.

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(z)(1) it is requested that the proposed alternative be authorized.

Document Control Desk Enclosure RC-17-0123 CR-16-01194 Page 10 of 10 7

PRA Quality Reference Attachment I Summary Statement of VCSNS Unit 1 PRA Model Capability for use in Risk-Informed Inservice Inspection Program Licensing Actions.

8 Duration of Proposed Alternative:

The alternative will be used at VCSNS until the end of the Fourth Ten-Year ISI Interval, subject to the review and update guidance of NEI 04-05. The Fourth Ten-Year ISI Interval is currently scheduled to end on December 31, 2023.

9 Precedents:

The proposed alternative in this 10CFR50.55a Request was included in a Third Interval Relief Request for VCSNS. This Relief Request was submitted to the NRC for approval per 10CFR50.55a(a)(3)(i) in VCSNS Letter No. RC-04-0148, dated September 8, 2004. Based upon the information provided in the RI-ISI template, the request to implement the RI-ISI methodology on Class 1 and 2 piping welds was approved by the NRC for VCSNS's Third Ten-Year ISI Interval in letter dated September 6, 2005 (TAC No. MC4323)

(ML052300616).

10

References:

1. ASME Code Section XI, Division 1, 2007 Edition through 2008 Addenda

2. Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk Informed Inservice Inspection Procedure.

3. ASME Code Case N-578-1, Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B Section XI, Division 1.

Document Control Desk RC-17-0123 CR-16-01194 Page 1 of 13 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT 1 Summary Statement of VCSNS Unit 1 PRA Model Capability for use in Risk-Informed Inservice Inspection Program Licensing Actions.

Document Control Desk RC-17-0123 CR-16-01194 Page 2 of 13

SUMMARY

STATEMENT OF VCSNS UNIT 1 PRA MODEL CAPABILITY FOR USE IN RISK-INFORMED INSERVICE INSPECTION PROGRAM LICENSING ACTIONS Introduction South Carolina Electric and Gas (SCE&G) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA model. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the VCSNS Unit 1 PRA.

PRA Maintenance and Update The SCE&G risk management process ensures that the PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in VCSNS procedure NL-126 (Probabilistic Risk Assessment Activities). This procedure delineates the responsibilities and guidelines for updating the full power internal events PRA models at VC Summer. It also defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following activities are routinely performed:

Design changes and procedure changes are reviewed for their impact on the PRA model.

New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

Maintenance unavailabilities are captured, and their impact on CDF is trended.

Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated every 2 or 3 years.

In addition to these activities, VCSNS risk PRA guidelines provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes:

Documentation of the PRA model, PRA products, and bases documents.

The approach for controlling electronic storage of PRA update information, PRA models, and PRA applications.

Guidelines for updating the full power, internal events PRA model.

Document Control Desk RC-17-0123 CR-16-01194 Page 3 of 13 Guidance for use of quantitative and qualitative risk models in support of the Equipment Out Of Service (EOOS) risk monitor for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10CFR50.65 (a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately 18 month cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant.

VCSNS performed a regularly scheduled update to the Unit 1 PRA model in 2016.

PRA Self Assessment and Peer Review Several assessments of technical capability have been made, for the VCSNS Unit 1 PRA model. These assessments are as follows:

Independent PRA peer reviews were conducted under the auspices of the Pressurized Water Reactor Owners Group (PWROG)\\ following the Industry PRA Peer Review process [1] in 2002 and 2016. These peer reviews included assessments of the PRA model maintenance and update process.

Self-assessments of the VCSNS PRA were conducted in October 2007 and November 2011 to the versions of RG1.200 that were current at those times. In November 2011, the VCSNS PRA model was also assessed to the ASME/ANS standard for PRA quality. Some gaps with respect to Westinghouse guidance (WCAPs) were identified during these reviews. These gaps have since been corrected.

A summary of the disposition of 2002 Industry PRA Peer Review A and B level facts and observations (F&Os) for the VCSNS model was documented via the corrective action program (CR-02-03871).

A PRA model update was completed in 2016 resulting in the revision 8a updated model.

In updating, changes were made to the PRA to address gaps identified in the November 2011 self-assessment. Following the update, an industry (PWROG) full scope peer review was performed. F&Os associated with the 2016 peer review have not yet been resolved. The following discussion describes how the F&Os relate to RI-ISI.

Assessment of PRA Capability Needed for Risk-Informed Inservice Inspection EPRI report TR01021467 (Reference 11) provides guidance on the ASME PRA standard capability category for each Supporting Requirement that is needed to support

Document Control Desk RC-17-0123 CR-16-01194 Page 4 of 13 RI-ISI. Some of the ASME standard Supporting Requirements listed in the EPRI report were identified as Not Met in the 2016 Peer Review.

The disposition of Not Met ASME Supporting Requirements required by EPRI TR019121467 is shown in Table 1 below:

Document Control Desk RC-17-0123 CR-16-01194 Page 5 of 13 Table 1: STATUS OF OPEN GAPS TO THE ASME PRA STANDARD SR F&O Description Impact Discussion IE-A5/A6 A systematic evaluation of each system, including support systems, to assess the possibility of an initiating event occurring due to a failure of the system could not be found.

(This F&O originated from SR IE-A5.)

Likely impact on RI-ISI:

None.

The initiating event list in the VCSNS PRA was based on a review of other risk assessments, plant operating history and plant design. This included a review of support systems. This is documented in calculation DC00300-013 Rev.

1. The initiator list stood unchallenged from 1993 to 2016. It is unlikely that another review of systems will change the initiating event list.

IE-A8 Documentation that plant personnel (e.g.,

operations, maintenance, engineering, safety analysis) have been interviewed to determine if potential initiating events have been overlooked could not be located.

(This F&O originated from SR IE-A8)

Likely impact on RI-ISI:

None.

No VC summer personnel have challenged the completeness of the initiating event list in the VC summer PRA from 1993 to the time of the peer review in 2016. It is unlikely that interviews will result in identification of overlooked initiating events.

IE-C1 CN-RAM-14-030, Appendix A documents the quantification of initiators. Both plant-specific and generic data are used in the development of the initiator mean values and uncertainties.

However, the use of more recent generic data is required to help assure that the failure rates are reflective of current industry experience. (this F&O went on to suggest using NUREG/CR-6928)

Also, the approach documented in CN-RAM 030, section 5 not to update the LOCA initiators because that would have a significant impact on CDF is not technically valid.

Likely impact on RI-ISI:

None.

A sensitivity study was performed by using the latest (2015) NUREG/CR-6928 initiating event frequencies. The only change in consequence keys used is that SI cold leg injection pathways (through XVC-8998A, B and C) went from MEDIUM in the 8b model to HIGH in the initiator sensitivity study. The increase is due to an increase of almost a factor of 4 for medium LOCA. Since there have been no new medium LOCA events in the industry, this difference could not be based on new data. The difference may be based on either binning or expert elicitation. The data set and breaks size binning used for the current VCSNS PRA model

Document Control Desk RC-17-0123 CR-16-01194 Page 6 of 13 Table 1: STATUS OF OPEN GAPS TO THE ASME PRA STANDARD SR F&O Description Impact Discussion (This F&O originated from SR IE-C1) is considered sufficient for medium LOCA. The F&O does not impact RI-ISI.

IE-C2 A review of recent LERs and plant initiator precursor events could not be found. The initiating events analysis is based on DC00300-150, Rev. 0, which only considered LERs through 3/22/08.

(This F&O originated from SR IE-C2)

Likely impact on RI-ISI:

Small impact or no impact.

A sensitivity study was performed by using the latest (2015) NUREG/CR-6928 initiating event frequencies. NUREG/CR-6928 used a review of LERs for applicable initiating event frequencies.

The sensitivity study showed that no consequence key ranking changed as a result of new data (see answer for IE-C1 above).

IE-C5 CN-RAM-14-030 section 8.1 indicates that an industry availability factor was applied in the VCS initiator quantifications.

(This F&O originated from SR IE-C5.)

Likely impact on RI-ISI:

None.

CN-RAM-14-030 uses DC-00300-150 which Baysean updates generic data with plant specific failures and critical hours over the calendar year period used.

Therefore, plant specific availability is included and this concern has already been addressed for the VCSNS PRA.

Document Control Desk RC-17-0123 CR-16-01194 Page 7 of 13 Table 1: STATUS OF OPEN GAPS TO THE ASME PRA STANDARD SR F&O Description Impact Discussion SY-A4 No system design walkdowns were found to support recent modeling updates.

(This F&O originated from SR SY-A4)

Likely impact on RI-ISI:

None.

Walkdowns of recent system modifications have been done in support of Fire PRA human reliability. These are documented for fire PRA.

They apply to the same systems (such as alternate seal injection) that support the internal events PRA. No modeling issues were noted during the walkdowns.

HR-G7 The notes in Table 1 of DC0300-149 include assessments that are inadequate alone to determine level of dependence:

1-Non-quantitative time intervals are used in many cases to define dependence. This should be based on the timing developed in the HRA.

2-Should be separated in time etc.

(This F&O originated from SR QU-C2)

Likely impact on RI-ISI:

Small.

A sensitivity study was performed using the HRA calculator dependent events method and no consequence keys increased in significance. Therefore, resolution of this F&O is not expected to add significance to piping segments.

DA-B2 No consideration for outliers considered. No distinction made between valves that are frequently operated and those infrequently operated. SR mapping claims this is N/A.

(This F&O originated from SR DA-B2)

Likely impact on RI-ISI:

None.

PRA failure rates are per hour or per demand.

Per demand failure rates already take into account frequency of manipulation. Valves that are seldom checked have a longer mission time for transfer. Identifying seldom operated valves as "outliers" will not change reliability values.

DA-C5 Consideration of repeated failures within a short time interval were only noted for EDGs in the review of the data analysis.

(This F&O originated from SR DA-C5)

Likely impact on RI-ISI:

None.

The F&O is concerned with over-counting failures by counting repeated failures over a short period of time separately. First, the VCSNS data analysis has not identified repeated failures in a short period in the data.

Second, if this had happened the calculated failure rate would be conservatively high.

Document Control Desk RC-17-0123 CR-16-01194 Page 8 of 13 Table 1: STATUS OF OPEN GAPS TO THE ASME PRA STANDARD SR F&O Description Impact Discussion Therefore, the current evaluation, with slightly higher failure rate would be conservative.

DA-C6 There is no evidence that additional demands from post-maintenance testing were considered.

(This F&O originated from SR DA-C6)

Likely impact on RI-ISI:

None.

This is a documentation issue. If more demands were counted, this would lower the failure rate and tend to cause segments to be less significant.

IFEV-A6 The initiating event frequency was based on generic information.

(This F&O originated from SR IFEV-A6)

Likely impact on RI-ISI:

Small and will not add significance.

Since there is no history of flooding at VCSNS any Bayesian updating with plant-specific data would cause the flood frequencies to be slightly lower because of updating with zero events.

The frequencies being used are slightly conservative.

IFEV-A7 The documentation excluded human induced flooding based on an assumption that cannot happen at power due to limited maintenance activities. This is inconsistent with the EPRI guidance.

Likely impact on RI-ISI:

Small and will not add significance.

Limited on-line maintenance makes human induced flooding less significant and it should not affect the failure probability or consequence for any pipe welds.

IFQU-A6 Impact of flooding on internal events HEPs was considered but it appears that the effect on HEPs was performed with not sufficient basis.

For example, from notebook CN-RAM-13-048,

""For Tdelay 10 minutes, an increase of two minutes was added to Texe and Tcog in the time window. For Tdelay > 10 minutes, an Likely impact on RI-ISI, Small and not likely to add significance.

Operator interviews were conducted and documented in CN-RAM-13-048 for time estimates. The HRA times were conservative relative to the operator interviews to account for potential distractors and operator stress level resulting from the flood. A more detailed approach would not add significance for the RI-ISI.

Document Control Desk RC-17-0123 CR-16-01194 Page 9 of 13 Table 1: STATUS OF OPEN GAPS TO THE ASME PRA STANDARD SR F&O Description Impact Discussion increase of five minutes was added to Texe and Tcog."" This appears to have no technical basis.

QU-D2/AS-A5 Cutsets and sequences have been reviewed to determine whether they accurately reflect plant design and operation. However, based on the overly conservative assumptions that go into the constituent analyses, some of the dominant cutsets are judged not to be realistic.

(This F&O originated from SR QU-D2)

Likely impact on RI-ISI:

Small and will not add significance to any segment.

This F&O is about applying recovery credit for systems that already have significant redundancy. In each case cited (in F&O AS-A5) the claim is that uncredited recovery options are available. Since the claim is of over-conservatism, resolution of this F&O will not make segments more significant. Also, since the systems in question already have redundancy, the effect of resolution is expected to be small.

Document Control Desk RC-17-0123 CR-16-01194 Page 10 of 13 In the risk-informed inservice inspection (RI-ISI) program at VCSNS, the PWROG RI-ISI methodology is used to define alternative inservice inspection requirements. Plant-specific PRA derived risk significance information is used during the RI-ISI plan development to support the consequence assessment, risk ranking and delta risk evaluation steps.

The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by two processes in the EPRI methodology.

First, PRA consequence results are binned into one of three conditional core damage probability (CCDP) and conditional large early release probability (CLERP) ranges before any welds are chosen for RI-ISI inspection. Table 2 illustrates the binning process.

Table 2 Consequence Results Binning Groups Consequence Category CCDP Range CLERP Range High CCDP > 1E-4 CLERP > 1E-5 Medium 1E-6 < CCDP 1E-4 1E-7 < CLERP 1E-5 Low CCDP 1E-6 CLERP 1E-7 The risk importance of a weld is therefore not tied directly to a specific PRA result.

Instead, it depends only on the range in which the PRA result falls. The wide binning provided in the methodology generally reduces the significance of specific PRA results.

Secondly, the influence of specific PRA consequence results is further reduced by the joint consideration of the weld failure potential via a non-PRA-dependent damage mechanism assessment. The results of the consequence assessment and the damage mechanism assessment are combined to determine the risk ranking of each pipe segment (and ultimately each element) according to the EPRI Risk Matrix. The Risk Matrix, which equally takes both assessments into consideration, is reproduced below.

Document Control Desk RC-17-0123 CR-16-01194 Page 11 of 13 POTENTIAL FOR PIPE RUPTURE PER DEGRADATION MECHANISM SCREENING CRITERIA CONSEQUENCES OF PIPE RUPTURE IMPACTS ON CONDITIONAL CORE DAMAGE PROBABILITY AND LARGE EARLY RELEASE PROBABILITY NONE LOW MEDIUM HIGH HIGH FLOW ACCELERATED CORROSION LOW Category 7 MEDIUM Category 5 HIGH Category 3 HIGH Category 1 MEDIUM OTHER DEGRADATION MECHANISMS LOW Category 7 LOW Category 6 MEDIUM Category 5 HIGH Category 2 LOW NO DEGRADATION MECHANISMS LOW Category 7 LOW Category 7 LOW Category 6 MEDIUM Category 4 These facets of the methodology reduce the influence of specific PRA results on the final list of candidate welds.

The limited use of specific PRA results in the RI-ISI process is also reflected in the risk-informed license application guidance provided in Regulatory Guide 1.174 [10].

Section 2.2.6 of Regulatory Guide 1.174 provides the following insight into PRA capability requirements for this type of application:

There are, however, some applications that, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.

An example is risk-informed inservice inspection (RI-ISI). In this application, risk significance was used as one criterion for selecting pipe segments to be periodically examined for cracking. During the staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary.

Therefore, the staff review of plant-specific RI-ISI typically will include only a limited scope review of PRA technical acceptability.

Further, Table 1.3-1 of the ASME PRA Standard(1) identifies the bases for PRA capability categories. The bases for Capability Category I for scope and level of detail attributes of the PRA states:

Resolution and specificity sufficient to identify the relative importance of the contributors at the system or train level including associated human actions.

Based on the above, in general, Capability Category I should be sufficient for PRA quality for a RI-ISI application.

The EPRI methodology further provides an alternate means to estimate the pipe rupture consequence, namely lookup tables. By using lookup tables, PRA analysis is not involved, and the impact of the loss of systems or trains is done in a generic (not plant-

Document Control Desk RC-17-0123 CR-16-01194 Page 12 of 13 specific) fashion. This allowable alternative underscores the relatively low dependence of the process on specific PRA capabilities.

In addition to the above, it is noted that welds are not eliminated from the ISI program on the basis of risk information. The risk significance of a weld may fall from Medium Risk Ranking to Low Risk Ranking, resulting in it not being a candidate for inspection.

However, it remains in the program, and if, in the future, the assessment of its ranking changes (either by damage mechanism or PRA risk) then it can again become a candidate for inspection. If a weld is determined, outside the PRA evaluation, to be susceptible to either flow-accelerated corrosion (FAC), inter-granular stress corrosion cracking (IGSCC) or microbiological induced cracking (MIC) in the absence of any other damage mechanism, then it moves into an "augmented" program where it is monitored for those special damage mechanisms. That occurs no matter what the Risk Ranking of the weld is determined to be.

1 Table A-1 of Regulatory Guide 1.200 identifies the NRC staff position as "No objection" to Section 1.3 of the ASME PRA Standard, which contains Table 1.3-1.

Conclusion Regarding PRA Capability for Risk-Informed ISI The VCSNS PRA models continue to be suitable for use in the RI-ISI application. This conclusion is based on:

the PRA maintenance and update processes in place, the PRA technical capability evaluations that have been performed and are being planned, and the RI-ISI process considerations, as noted above, that demonstrate the relatively limited reliance of the process on PRA capability.

References

1. NEI-00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Rev. A3.

2. U.S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 2.

3. VC Summer Nuclear Station Probabilistic Risk Assessment Peer Review Final Report, Westinghouse, December 2002.

4. LTR-RRA-05-53, Summary of V.C. Summer PRA Regulatory Guide 1.200 App. B/

ASME PRA Standard Deltas Assessment, November, 2005

Document Control Desk RC-17-0123 CR-16-01194 Page 13 of 13

5. PRA Model PRA Standards Compliance Assessment Dates, November 7th-November 11th, 2011 SA 11-NL-06

6. PWROG-160511-P, Rev. 0, Peer Review of the V. C. Summer Nuclear Station Internal Events and Internal Flood Probabilistic Risk Assessment, February, 2017

7. Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant ASME RA-Sa-2009, New York, New York, December 2005.

8.

U.S. Nuclear Regulatory Commission Memorandum to Michael T. Lesar from Farouk Eltawila, "Notice of Clarification to Revision 1 of Regulatory Guide 1.200,"

for publication as a Federal Register Notice, July 27, 2007.

9.

Revised Risk-Informed Inservice Inspection Evaluation Procedure, EPRI TR-112657, Revision BA, December 1999.

10. U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 1, November 2002.

11. Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs, EPRI-TR01021467

Document Control Desk RC-17-0123 CR-16-01194 Page 1 of 3 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT 2 VCSNS - Inspection Location Selection Comparison Between Previously Approved and Revised RI-ISI Program by Risk Category

Document Control Desk RC-17-0123 CR-16-01194 Page 2 of 3 VCSNS - Inspection Location Selection Comparison Between Previously Approved and Revised RI-ISI Program by Risk Category System Risk Consequence Rank Failure Potential Previously Approved (Third Interval)

Updated (Fourth Interval)

Category Rank DMs Rank Weld Count(1)

RI-ISI(1)

Weld Count(1)

RI-ISI(1) 01RC 2

High High TASCS, TT Medium 23 9

23 9

01RC 2

High High TASCS Medium 2

0 2

0 01RC 2

High High TT Medium 36 7

36 7

02RHR 2

High High TASCS Medium 5

2 5

2 03SI 2

High High TASCS, TT Medium 7

4 7

4 03SI 2

High High TASCS Medium 13 1

13 1

03SI 2

High High TT Medium 8

2 8

2 04CS 2

High High TASCS, TT Medium 7

1 7

1 04CS 2

High High TT Medium 18 6

18 6

08EF 2

High High TT Medium 3

1 01RC 4

Medium High None Low 166 18 166 19 01RC 4 (2)

Medium (High)

High None (PWSCC)

Low (Medium) 16 1(2) 16 0(2) 02RHR 4

Medium High None Low 53 6

53 6

03SI 4

Medium High None Low 215 22 215 22 04CS 4

Medium High None Low 11 2

43 5

05SP 4

Medium High None Low 7

1 7

1 08EF 4

Medium High None Low 5

1 06MS 4

Medium High None Low 35 4

02RHR 5a Medium Medium TASCS Medium 6

1 6

1 03SI 5a Medium Medium TT, IGSCC Medium 4

1 4

1 03SI 5a Medium Medium TT Medium 10 0

10 0

03SI 5a Medium Medium IGSCC Medium 16 2

16 2

04CS 5a Medium Medium TT Medium 4

1 2

1 07FW 5a Medium Medium TASCS Medium 7

1 7

1 08EF 5a Medium Medium TT Medium 3

2(3)

Document Control Desk RC-17-0123 CR-16-01194 Page 3 of 3 (Continued)

VCSNS - Inspection Location Selection Comparison Between Previously Approved and Revised RI-ISI Program by Risk Category System Risk Consequence Rank Failure Potential Previously Approved (Third Interval)

Updated (Fourth Interval)

Category Rank DMs Rank Weld Count(1)

RI-ISI(1)

Weld Count(1)

RI-ISI(1) 01RC 6a Low Medium None Low 7

0 02RHR 6a Low Medium None Low 237 0

237 0

03SI 6a Low Medium None Low 548 0

411 0

04CS 6a Low Medium None Low 401 0

35 0

05SP 6a Low Medium None Low 226 0

226 0

06MS 6a Low Medium None Low 143 0

108 0

07FW 6a (3)

Low (High)

Medium None (FAC)

Low (High) 15 0

15 0

07FW 6a Low Medium None Low 52 0

52 0

08EF 6a Low Medium None Low 75 0

80 0

09SW 6a (5a)

Low (Medium)

Medium None (MIC, PIT)

Low (Medium) 35 0

04CS 6b Low Low TT Medium 2

0 01RC 7a Low Low None Low 7

0 03SI 7a Low Low None Low 4

0 141 0

04CS 7a Low Low None Low 334 0

05SP 7a Low Low None Low 14 0

14 0

09SW 7a (6b)

Low (Low)

Low None (MIC, PIT)

Low (Medium) 35 0

Total 2399 90 2399 97 Notes:

1. A dash shown under Weld Count or RI-ISI indicates that due to changes in Risk Rankings, there were no welds for that particular listing in the given interval.

2. In accordance with 10CFR50.55a(g)(6)(ii)(F), welds subject to PWSCC were selected for examination per Code Case N-770-1 during the Third Interval and examined under that program. Welds for which no other degradation mechanism has been postulated, were examined solely under the Code Case N-770-1 Program and were removed from consideration during the RI-ISI element selection process. During the Fourth Interval this same process will be followed using the version of Code Case N-770 that is referenced in the published version of 10CFR50.55a.

3. A second Class 2 weld in the EF system, Risk Category 5a was selected to supplement a Third Interval selection which had limited coverage.