RC-14-0002, Response to Request for Additional Information Regarding License Amendment to Adopt NFPA 805

Response to Request for Additional Information Regarding License Amendment to Adopt NFPA 805

ML14013A074
Person / Time
Site:	Summer
Issue date:	01/09/2014
From:	Gatlin T South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR-06-00055, RC-14-0002
Download: ML14013A074 (6)

Text

Thomas D. Gatlin Vice President,Nuclear Operations 803.345.4342 January 9, 2014 A SCANA COMPANY RC-14-0002 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 Attn: S. A. Williams

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT REQUEST- LAR-06-00055 LICENSE AMENDMENT REQUEST TO ADOPT NFPA 805 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

References:

1. Thomas D. Gatlin, SCE&G, Letter to NRC Document Control Desk, License Amendment Request - LAR-06-00055, "License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)" dated November 15, 2011 (RC-11-0149) [ML11321A172]

2. Shawn A. Williams, NRC, to Thomas D. Gatlin "Virgil C. Summer Nuclear Station Unit - 1 (VCSNS) - Request for Additional Information (TAC NO. ME7586)" dated November 7, 2013 [ML13308B800]

3. Thomas D. Gatlin, SCE&G, Letter to NRC Document Control Desk, License Amendment Request - LAR-06-00055, "License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information" dated October 10, 2012 (RC-12-0142)

4. Thomas D. Gatlin, SCE&G, Letter to NRC Document Control Desk, License Amendment Request - LAR-06-00055, "License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information" dated October 14, 2013 (RC-13-0142)

South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for South Carolina Public Service Authority pursuant to 10 CFR 50.90, submitted a License Amendment Request (LAR) per Reference 1 to adopt NFPA 805. NRC review and audit of this request determined that additional information was required and a Request for Virgil C.Summer Station

• Post Office Box 88 - Jenkinsville, SC- 29065

• F (803) 941-9776 VOD9

• 'k

Document Control Desk LAR-06-00055 RC-14-0002 Page 2 of 2 Additional Information (RAI) was issued per Reference 2. Attachment 1 of this letter provides SCE&G's response to the RAIs.

If you have any questions about this submittal, please contact Mr. Bruce L. Thompson at (803) 931-5042.

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

Executed on Thomas D. Gatlin RLP/TDG/ts

Attachment:

1. Probabilistic Risk Assessment (PRA) Request for Additional Information (RAI)

Responses c: K. B. Marsh S. A. Byrne J. B. Archie N. S. Carns J. H. Hamilton J. W. Williams W. M. Cherry V. M. McCree S. A. Williams NRC Resident Inspector S. E. Jenkins Paulette Ledbetter K. M. Sutton NSRC RTS (CR-06-00055)

File (813.20)

PRSF (RC-14-0002)

Document Control Desk Attachment LAR-06-00055 RC-14-0002 Page 1 of 4 VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ATTACHMENT Probabilistic Risk Assessment (PRA)

Request for Additional Information (RAI) Responses

Document Control Desk Attachment LAR-06-00055 RC-14-0002 Page 2 of 4 Probabilistic Risk Assessment (PRA) RAI 10.02 In a letter dated October 10, 2012 (Agencywide Document Access and Management System (ADAMS) Accession No. ML2297A218), the licensee responded to PRA RAI 10 and provided a detailed description of a series of operator actions to shutdown the facility using a preferred success path after main control room (MCR) abandonment. It appears from the response that this single human error probability (HEP) value is used for every MCR abandonment scenario.

a. Describe whether any values other -than 0.05 are used to characterize the HEP following MCR abandonment and whether the 0.05 value also represents the conditional core damage probability (CCDP).

b. If any values other than 0.05 are used, e.g., 1.0, provide the other values, a characterization of the scenarios that these values are used, and a summary of how each value is developed. This information should include explanations of how the following scenarios are addressed.

i. Scenarios where the fire fails few functions aside from MCR habitability and successful shutdown is straightforward.

ii. Scenarios where the fire could cause some recoverable functional failures or spurious operations that complicate the shutdown but successful shutdown is likely.

iii. Scenarios where the fire induced failures cause great difficulty for shutdown by failing multiple functions andlor complex spurious operations that make successful shutdown unlikely.

c. If no values other than 0.05 are used, please explain how scenarios characterized under 10.02.b.i, 10.02.b.ii, and 10.02.b.iii are included in your MCR abandonment evaluations.

SCE&G Response At the time of the LAR submittal, the Human Reliability Analysis (HRA) for abandonment was based on the best available information at the time regarding the MCR abandonment procedure, with the knowledge that the procedure would be changing with the change of shutdown strategy away from the Appendix R Self Induced Station Blackout (SISBO) strategy. As discussed in the public meeting on June 20, 2013 (ADAMS Accession No. ML1317OA1 30), SCE&G is developing new procedures associated with control room abandonment. These new procedures are being evaluated for feasibility, and an updated MCR abandonment HRA and Fire Probabilistic Risk Assessment (FPRA) modeling is underway. The new modeling uses a single HEP

Document Control Desk Attachment LAR-06-00055 RC-14-0002 Page 3 of 4 value. It is part of an "OR" gate that also includes failures of instrumentation needed by the operator as a cue or indication in the MCR abandonment procedure as well as failures of equipment that results in loss of function. The effect of the updated analysis will be incorporated in the new results along with other model updates, and will be reported in a revised Attachment G and revised Attachment W. SCE&G will submit the revised LAR Attachment G "Recovery Actions Transition" along with a revised LAR Attachment W "Fire PRA Insights" by February 25, 2014.

PRA RAI 96 Regarding Fact and Observation (F&O) CF-AI-01, the licensee applies Option #2 of Task 10 of NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," for circuit analysis, as recommended by the peer review. However, by NRC memorandum, "Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis," dated June 14, 2013 (ADAMS Accession No. ML13165A209), Option #2 "does not provide an adequate method for quantifying the likelihood of hot short-induced spurious operations." As a result, provide a sensitivity analysis, evaluating core damage frequency (CDF),

large early release frequency (LERF), Delta (A)CDF, and ALERF, by substituting an acceptable method for Option #2. It should be noted that Option #1 of Task 10 of NUREG/CR-6850, absent control power transformer (CPT) credit, is an acceptable method, as well as the cited staff guidance NUREGICR-6850.

SCE&G Response Circuit Failure Likelihood in the V.C. Summer Fire PRA is based on use of Option 1. As noted in response PRA RAI 09 in Reference 3 dated October 10, 2012; credit for CPT was removed and sensitivity results presented. Option 2 was not used to determine Circuit Failure Likelihood in the V.C. Summer Fire PRA. Because Option 1 without CPT credit is an acceptable method, no additional sensitivity analysis is required.

Document Control Desk Attachment LAR-06-00055 RC-14-0002 Page 4 of 4 PRA RAI 97 Attachment S, Table S-1, Item ECR50799, of the License Amendment Request (LAR), (ADAMS Accession No. ML11332A076) describes a proposed change of providing lower leakage Reactor Coolant Pump (RCP) seals and crediting them in the Fire Probabilistic Risk Assessment (FPRA). The exact RCP seal package is not specified in the LAR; however, the licensee indicated by teleconference that earlier plans to install the Westinghouse passive thermal shutdown seal (SHIELD) had been abandoned for a different RCP seal package.

As a result:

a. Describe the type of seal package modeled in the PRA which supports the risk estimates reported in the November 15, 2011, LAR.

b. If a different seal package is planned to be installed to support transition to NFPA 805 from that which was used in the PRA supporting the November 15, 2011, LAR, identify the new seal package. Also, please answer the following:

i. Provide the revised risk estimates of the impacts, including the basis for any revised assumptions.

ii. If the new seal package model causes the total change in risk from transitioning to NFPA 805 to exceed the acceptance guidelines of RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," describe the plans to demonstrate that the risk increase is acceptable.

iii. Discuss whether any currently proposed license conditions are affected by these changes.

SCE&G Response The seal package modeled in the PRA supporting the November 15, 2011 LAR is based on the Flowserve N9000 seals. VCSNS intends to install the Flowserve N9000 seal package.

As noted in Reference 4 LAR Attachment S Table S-1, seal replacement will be completed by the end of 2015. Reference 4 LAR Attachment S Table S-2, item 22 also requires update of the PRA to reflect the as-built modifications.