RC-18-0091, License Amendment Request - LAR-16-01490 National Fire Protection Association Standard 805 Program Revisions

License Amendment Request - LAR-16-01490 National Fire Protection Association Standard 805 Program Revisions

ML18242A658
Person / Time
Site:	Summer
Issue date:	08/29/2018
From:	Lippard G South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML18242A657	List: RC-18-0091, License Amendment Request - LAR-16-01490 National Fire Protection Association Standard 805 Program Revisions
References
RC-18-0091
Download: ML18242A658 (103)

Text

Sensitive Information -Withhold From Public Disclosure Under 10 CFR 2.390 George A. Lippard

< sCE&G@*

Vice President, Nuclear Operations 803.345.4810 A SCANA COMPANY August 29, 2018 RC-18-0091 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT REQUEST-LAR-16-01490 NATIONAL FIRE PROTECTION ASSOCIATION (NFPA) STANDARD 805 PROGRAM REVISIONS

References:

1. Shawn Williams, NRC, Letter to Thomas Gatlin, "Virgil C. Summer Nuclear Station,. Unit 1 - Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (TAC No. ME7586)", dated February 11, 2015. (ML14287A289)

2. Public Meeting March 15, 2018 Pre-Application Meeting with South Carolina Electric & Gas Company (SCE&G) to discuss a proposed license amendment regarding NFPA 805 for V. C. Summer, Unit 1, February 23, 2018.

(ML 18054AOOO)

3. Anne Boland, NRC, Letter to Michael Tschiltz, "Recommended Content for License Amendment Requests that Seek Changes to License Conditions That Were Established in Amendments to Adopt National Fire Protection Association Standard 805 But Have Yet to be Fully Implemented", dated March 2, 2016.

(ML16015A416)

Pursuant to 10 CFR 50.90, South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for South Carolina Public Service Authority, hereby requests an amendment to the Virgil C. Summer Nuclear Station (VCSNS) Unit 1 Renewed Facility Operating License No.

NPF-12. SCE&G has adopted the NFPA 805 program as approved by the NRC in the letter dated February 11, 2015 (Reference 1 ). The station proposes to conduct changes to the license conditions that were discussed with the NRC in a Public Meeting on March 15, 2018 (Reference 2). SCE&G requests approval of the following proposed changes:

Changes to plant modifications evaluated using Fire PRA Methods and approaches that have been previously accepted in Reference 1 or that have been accepted for another station by applying the guidance in Reference 3 following option B.

Request for Approval of Performance-Based Alternatives for Chapter 3 NFPA 805 (10 CFR 50.48(c)(2)(vii)).

o NFPA 805 Section 3.3.4 Insulation Materials o

NFPA 805 Section 3.3.5.1 Wiring above Suspended Ceilings, Attachments 5, 6, and 7 of transmitted herewith contains sensitive information.

When separated from these enclosures, this transmittal document is decontrolled.

V. C Summer Nu dear Station* P. 0. Box 88

• Jenkinsville, South Carolina* 29065

• F (803) 941-9776

• www.sceg.com

(without attachment)

Document Control Desk LAR-16-01490 RC-18-0091 Page 1 of 18 VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ENCLOSURE 1 LICENSE AMENDMENT REQUEST TO REVISE NFPA 805 PROGRAM

Document Control Desk LAR-16-01490 RC-18-0091 Page 2 of 18 ENCLOSURE 1 1.0

SUMMARY

DESCRIPTION............................................................................................. 3 1.1 Current Licensing Basis....................................................................................... 3 1.2 Issues Identified During Transition Period............................................................ 4 1.3 Reasons for the Proposed Changes.................................................................... 5 1.3.1 Proposed Modification Scope Changes................................................... 5 1.3.2 Proposed Fundamental Fire Protection Program and Design Elements.. 6 2.0 DETAILED DESCRIPTION............................................................................................. 7 2.1 Proposed Modification Scope Changes............................................................... 7 2.2 Proposed Fundamental Fire Protection Program and Design Elements............... 7

3.0 TECHNICAL EVALUATION

............................................................................................ 8 3.1 Proposed Modification Scope Changes............................................................... 8 3.2 Proposed Fundamental Fire Protection Program and Design Elements............... 8

4.0 REGULATORY EVALUATION

....................................................................................... 8 4.1 Regulatory Summary of Modification Changes.................................................... 8 4.2 Regulatory Summary of Changes to Fundamental Fire Protection Program and Design Elements.................................................................................................. 9 4.3 Precedent.......................................................................................................... 10 4.3.1 Precedence for Proposed Modification Scope Changes.........................10 4.3.2 Precedence for Proposed Changes to the Fundamental Fire Protection Program and Design Elements...............................................................10 4.4 No Significant Hazards Consideration Determination......................................... 11 4.5 Conclusions....................................................................................................... 13

5.0 ENVIRONMENTAL CONSIDERATION

......................................................................... 13

6.0 REFERENCES

.............................................................................................................. 14 7.0 ACRONYMS................................................................................................................. 16 8.0 ATTACHMENTS........................................................................................................... 18

Document Control Desk LAR-16-01490 RC-18-0091 Page 3 of 18 1.0

SUMMARY

DESCRIPTION South Carolina Electric and Gas (SCE&G) has adopted the National Fire Protection Association (NFPA) 805 program for V. C. Summer Nuclear Station (VCSNS) as approved by Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) in letter dated February 11, 2015 (Reference 1). The station proposes to conduct changes to the license conditions that were discussed with the NRC in a Public Meeting on March 15, 2018 (Reference 2). The proposed changes include:

Changes to Plant Modifications This submittal follows the guidance established by NRC Letter to Michael Tschiltz, (Reference 3), specifically Option B. The proposed changes to plant modifications using Fire Probabilistic Risk Assessment (PRA) Methods and approaches have been accepted in a final safety evaluation for another station. Although the Fire PRA used in support of this License Amendment Request (LAR) primarily uses approaches approved in the NRC Safety Evaluation, the changes to the plant modifications include changes to PRA approaches that were not approved for use during the initial approval of the program:

o Updates for incipient detection system in Fire Areas CB06 and CB15 to reflect the latest modeling guidance as specified in NUREG-2180 (Reference 17).

o Updates to the generic ignition frequencies and manual non-suppression probabilities with the latest values reported in NUREG-2169 (Reference 18).

o Updates to reflect the heat release rate probability distributions and obstructed plume model analyses as described in NUREG-2178 (Reference 19).

These guidance documents were not approved for use during the original development and review of the VCSNS Fire PRA. The use of the latest industry guidance, either in the base model or as an implementation item, is reflected in more recent NFPA 805 SEs, including:

o Ginna - NFPA 805 SE dated June 25, 2018 (ML18114A025).

o Farley - SE dated October 17, 2016 (ML16232A000).

Fundamental Fire Protection Program and Design Elements VCSNS Request for Approval of Performance-Based Alternatives for Chapter 3 NFPA 805 under the provisions of 10 CFR 50.48(c)(2)(vii).

o NFPA 805 Section: 3.3.4 Insulation Materials is a new request based on the discovery of thermal insulation materials.

o NFPA 805 Section: 3.3.5.1 Wiring Above Suspended Ceiling - Revision to previously approved configuration to address additional areas that were discovered.

1.1 Current Licensing Basis The VCSNS Unit 1 Operating License (Reference 1) Condition 2.C.(18) requires SCE&G to implement and maintain a fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c). Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, VCSNS may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change

Document Control Desk LAR-16-01490 RC-18-0091 Page 4 of 18 does not require a change to a technical specification or a license condition, and the specific criteria listed in the license condition are satisfied.

One of the specific criterion of the license condition is 2.C.(18).c, Transition License Conditions. Section 2 of that license condition, states:

The licensee shall implement the modifications to its facility, as described in Attachment S, Table S-1, Plant Modifications Committed, of SCE&G letter RC-14-0196, dated December 11, 2014, by the end of the calendar year 2015. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

The original LAR-06-00055 (References 1 and 5) contained a list of committed plant modifications within Table S-1 and a list of Variances From Deterministic Requirements (VFDR) which are resolved by a plant modification within Attachment C. VFDRs are identified as Deterministic Requirement Open Item Description (DROID) at VCSNS.

During the NRCs review of the proposed program, the following changes were made to the original Table S-1 and Attachment C:

SCE&G letter RC-12-0142, dated October 10, 2012 (Reference 6)

Clarified the scope of two Engineering Change Requests (ECRs), ECR 50784 and ECR 50810, identified in LAR Table S-1 as committed plant modifications SCE&G letter RC-13-0142, dated October 14, 2013 (Reference 24)

SCE&G letter RC-14-0027, dated February 25, 2014 (Reference 8)

Updated the LAR Attachment C list of DROIDs which were resolved by a plant modification SCE&G letter RC-14-0067, dated May 2, 2014 (Reference 9)

Clarified changes made to Attachment C in RC-14-0027 SCE&G letter RC-14-0129, dated August 14, 2014 (Reference 7)

Clarified the scope of two Engineering Change Requests (ECRs), ECR 50784 and ECR 50810, identified in LAR Table S-1 as committed plant modifications SCE&G letter RC-14-0196, dated December 11, 2014 (Reference 10)

Provided the final LAR Table S-1, which was referenced in the NRC Safety Evaluation (Reference 1) as a Transition License Condition to complete the modifications listed in the table prior to the end of calendar year 2015 1.2 Issues Identified During Transition Period During the transition period the following issues were identified:

De-scoped Modifications:

o On March 24, 2016, the station determined that some aspects of ECR 50810 (NFPA 805 Hazards Mitigation) and ECR 50784 (NFPA 805 Circuit Protection) were de-scoped during the transition period. This issue was entered into the stations corrective action program as condition report CR-16-01490.

Document Control Desk LAR-16-01490 RC-18-0091 Page 5 of 18 o On October 20, 2016, the station identified additional committed modifications that were not implemented (CR-16-05291).

Fire PRA Issues o On March 31, 2016, the station identified that the calculated delta risk between compliant and variant plant did not meet the guidance of RG 1.174 (Reference 15). This issue was added to the stations corrective action program as CR-16-01602.

o Additional Fire PRA issues were identified in CR-16-00321, CR-16-01132, CR-16-04828, and CR-16-04829.

These issues were dispositioned as violations in NRC Triennial Fire Protection Inspection Report 05000395/2016010 (Reference 23).

In addition, during internal and external reviews of the fire protection program, the following two fundamental fire protection program elements were identified as requiring a license amendment:

NFPA 805 Section: 3.3.5.1 Wiring Above Suspended Ceiling (CR-18-01237)

NFPA 805 Section: 3.3.4 Insulation Materials (CR-16-03580) 1.3 Reasons for the Proposed Changes 1.3.1 Proposed Modification Scope Changes The completed scope of ECRs 50784 and 50810 is not consistent with the description of the ECRs in RC-12-0142 (ML12297A218) and RC-14-0129 (ML14227A737). SCE&G inappropriately used performance-based fire risk evaluations to de-scope modifications without obtaining the prior approval from the NRC (Reference 3).

The completion of the LAR Table S-1 modifications did not result in deterministic compliance for all the DROIDs dispositioned in RC-14-0027 (Reference 8) and RC-14-0067 (Reference 9). Fire Risk Evaluations were completed for DROIDs that were not brought into deterministic compliance by the completion of the Table S-1 modifications.

This event was added to the Corrective Action Program as CR-16-01490 and CR-16-05291. The inappropriate changes were identified and dispositioned as violations in NRC Triennial Fire Protection Inspection Report 05000395/2016010 (Reference 23).

SCE&G requests approval to modify the following second Transition License Condition found in VCSNS Operating License 2.C.(18), Fire Protection:

The licensee shall implement the modifications to its facility, as described in Attachment S, Table S-1, "Plant Modifications Committed," of SCE&G letter RC-14-0196, dated December 11, 2014, by the end of the calendar year 2015. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

Document Control Desk LAR-16-01490 RC-18-0091 Page 6 of 18 SCE&G proposes to revise the second Transition License Condition to revise the scope of ECR 50784 and ECR 50810 as follows:

The licensee shall implement the modifications to its facility, as described in Attachment S, Table S-1, "Plant Modifications Committed," of SCE&G letter RC-14-0196, dated December 11, 2014, as modified by, Attachments 2 and 3, of SCE&G letter RC-18-0091, dated August 29, 2018. The licensee shall maintain appropriate compensatory measures in place until issuance of the safety evaluation.

This change allows the DROIDs impacted by the modification scope change to be resolved using the performance-based fire risk evaluation methodology approved by the NRC.

1.3.2 Proposed Fundamental Fire Protection Program and Design Elements Reviews of the NFPA 805 implementation identified changes to the license bases that are required to address as-found conditions at the station. The station discovered two deviations from the NFPA 805 section 3.0, Fundamental Fire Protection Program and Design Elements on of which was previously approved under 10 CFR 50.48(c)(2)(vii). The proposed changes are:

NFPA 805 Section 3.3.5.1, Wiring Above Suspended Ceiling In the original LAR-16-00055 Request L2 (References 1 and 5) and as modified by a response to FPE RAI 08 by SCE&G letter RC-12-0142 (Reference 6) and FPE RAI 14.02 by SCE&G letter RC-13-0142 (Reference 24), VCSNS requested the NRC to review and approve a performance-based method. VCSNS demonstrated an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.5.1 for the existing wiring in suspended ceilings. See SE (Reference 1) Section 3.1.4.2 for the NRC staff's review of this request.

A revision is needed to the original 10 CFR 50.48(c)(2)(vii) Approval Request L2, due to the discovery of additional areas in the plant where wiring is installed above suspended ceilings. A full walkdown was completed to assess the extent of these installations, including the type of wiring, purpose of wiring, amount of wiring, presence of fire detection, and presence of fire suppression. Minor editorial corrections were also captured throughout the L2 approval request. This event was added to the Corrective Action Program as CR-18-01237.

NFPA 805 Section 3.3.4, Insulation Materials VCSNS has discovered that a new 10 CFR 50.48(c)(2)(vii) Approval Request L14 is needed for thermal insulation materials. The new request is based on the discovery of thermal insulation materials which do not meet the explicit requirements of NFPA 805, Section 3.3.4. A full walkdown was completed to assess the extent of these installations

Document Control Desk LAR-16-01490 RC-18-0091 Page 7 of 18 including the type of thermal insulation, purpose of thermal insulation, and amount of thermal insulation. This event was added to the Corrective Action Program as CR-16-03580.

2.0 DETAILED DESCRIPTION 2.1 Proposed Modification Scope Changes This submittal follows the guidance established by NRC Letter to Michael Tschiltz, (Reference 3), specifically Option B. This option evaluates proposed changes to plant modifications using Fire PRA Methods and approaches have been accepted in a final safety evaluation for another station. VCSNS de-scoped modifications for cables that were specified for protection as part of regulatory correspondence, but some identified modifications were eliminated and addressed via the fire risk evaluation process. The detailed description of the proposed modifications scope is provided in Enclosure 1, titled Plant Modifications Associated with the NFPA 805 Transition.

The Fire PRA used in support of this LAR primarily uses approaches reviewed as part of the NRC Safety Evaluation. However, certain approaches and guidance documents were not approved for use during the original development and review of the VCSNS Fire PRA. The use of the latest industry guidance is consistent with recent Safety Evaluations for Ginna (ML18114A025) and Farley (ML16232A000) NFPA 805 Programs.

The detailed description of the proposed change in PRA approaches is provided in this, Attachment 1 titled Plant Modifications Associated with the NFPA 805 Transition. The proposed updates to the VCSNS NFPA 805 Program includes:

Updates for incipient detection system in Fire Areas CB06 and CB15 to reflect the latest modeling guidance as specified in NUREG-2180.

Updates to the generic ignition frequencies and manual non-suppression probabilities with the latest values reported in NUREG-2169.

Updates to reflect the heat release rate probability distributions and obstructed plume model analyses as described in NUREG-2178.

2.2 Proposed Fundamental Fire Protection Program and Design Elements VCSNS Request for Approval of two Performance-Based Alternatives for NFPA 805 Chapter 3, Fundamental Fire Protection Program and Design Elements under the provisions of 10 CFR 50.48(c)(2)(vii). The first is a proposed revision to the original Approval Request L2, due to the discovery of additional areas in the plant where wiring is installed above suspended ceilings. The revision will also provide some minor editorial corrections that were captured throughout the L2 approval request. The second Performance-Based Alternatives is a new approval request L14 for thermal insulation materials. The new request is based on the discovery of thermal insulation materials which do not meet the explicit requirements of NFPA 805, Section 3.3.4, Insulation Materials. Enclosure 1, Attachment 9 titled NFPA 805 Chapter 3 Requirements for Approval 10 CFR 50.48(c)(2)(vii) contains the detailed description of the two requests for performance-based alternatives.

Document Control Desk LAR-16-01490 RC-18-0091 Page 8 of 18

3.0 TECHNICAL EVALUATION

3.1 Proposed Modification Scope Changes Fire Risk Evaluations were performed for each fire area that inappropriately de-scoped modifications and used performance-based fire risk evaluations. Enclosure 1,, Plant Modifications Associated with the NFPA 805 Transition contains a detailed discussion of the results for each evaluation. This submittal follows the guidance established by NRC Letter to Michael Tschiltz, (Reference 3), specifically Option B. In summary, the fire risk evaluations concluded that the change in risk is acceptable and safety margins are maintained. The three major elements of defense-in-depth have been maintained: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The new as-built risk results are provided in, Attachment 5 titled LAR Attachment W - Fire Risk Insights Updates.

3.2 Proposed Fundamental Fire Protection Program and Design Elements SCE&G evaluated the performance-based alternatives for the NFPA 805 section 3.3.4, Insulation Materials and NFPA 805 section 3.3.5.1, Wiring Above Suspended Ceilings.

The details are provided in Enclosure 1, Attachment 9 titled NFPA 805 Chapter 3 Requirements for Approval 10 CFR 50.48(c)(2)(vii). The Fire Protection Program engineering and administrative features and controls provide a level of risk management and performance that achieves the following criteria:

Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; Maintains safety margins; and Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

4.0 REGULATORY EVALUATION

4.1 Regulatory Summary of Modification Changes Regulatory Position C.3.1 of RG 1.205, Revision 1, (Reference 21), states that a license condition included in a NFPA 805 LAR should include: (1) a list of modifications being made to bring the plant into compliance with 10 CFR 50.48(c); (2) a schedule detailing when these modifications will be completed; and (3) a statement that the licensee shall maintain appropriate compensatory measures in place until implementation of the modifications are completed. VCSNS submitted a list of modifications and implementation items originally in LAR-06-00555 Attachment S, "Plant Modifications and Items to be Completed during Implementation." The updated LAR-16-00055 Attachment S (References 1 and 5) is provided in the letter from VCSNS to the NRC dated

Document Control Desk LAR-16-01490 RC-18-0091 Page 9 of 18 December 11, 2014 (Reference 10). VCSNS proposed amendment request supports the transition of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition), in that it:

Complies with the requirements in fire protection regulation 10 CFR 50.48(a),

10 CFR 50.48(c), and Is consistent with the guidance in:

o Regulatory Guide (RG) 1.205, Risk-Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants (Reference 21), and o NRC, Letter to Michael Tschiltz, Recommended Content of License Amendment Requests that Seek Changes to License Conditions That Were Established in Amendments to Adopt National Fire Protection Association Standard 805 But Have Yet to be Fully Implemented, dated March 2, 2016 (ML16015A416). [Except prior NRC approval for the de-scoped modifications was not obtained during the transition period.]

o NUREG-2180 (Reference 17) Updates for incipient detection system.

o NUREG-2169 (Reference 18) Generic ignition frequencies and manual non-suppression probabilities.

o NUREG-2178 (Reference 19) Heat release rate probability distributions and obstructed plume model analyses.

4.2 Regulatory Summary of Changes to Fundamental Fire Protection Program and Design Elements Section 3.1 of NFPA 805 specifically allows the use of alternatives to the NFPA 805, Chapter 3 fundamental fire protection program requirements that have been previously approved by the NRC. VCSNS used the systematic approach described in NEI 04-02, Revision 2 (Reference 25), as endorsed by the NRC in Regulatory Guide 1.205, Revision 1 (Reference 21), to assess the VCSNS fire protection program against the NFPA 805 Chapter 3 requirements. VCSNS requested approval for the use of performance-based methods to demonstrate compliance with fundamental fire protection program elements identified in the original LAR-06-00555 Attachment A Table B-1 (References 1 and 5). In accordance with 10 CFR 50.48(c)(2)(vii), VCSNS requested specific approvals be included in the license amendment approving the transition to NFPA 805. 10 CFR 50.48(c)(2)(vii) provides additional requirements related to NFPA 805 Chapter 3. 10 CFR 50.48(c)(2)(vii) states, in part:

(vii) Performance-based methods. Notwithstanding the prohibition in Section 3.1 against the use of performance-based methods, the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard. Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under § 50.90.

Document Control Desk LAR-16-01490 RC-18-0091 Page 10 of 18 In accordance with 10 CFR 50.48(c)(2)(vii) and the guidance provided in NEI 04-02 (Reference 25), SCE&G has included these requests for approval in Enclosure 1,.

NFPA 805 Section 3.3.5.1, Wiring Above Suspended Ceiling In the original LAR-06-00055 Request L2 (References 1 and 5) and as modified by a response to FPE RAI 08 by SCE&G letter RC-12-0142 (Reference 6),

VCSNS requested the NRC to review and approve a performance-based method. VCSNS demonstrated an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.5.1 for the existing wiring in suspended ceilings. See SE (Reference 1) Section 3.1.4.2 for the NRC staff's review of this request. A revision is needed to make corrections to the original 10 CFR 50.48(c)(2)(vii) Approval Request L2.

NFPA 805 Section 3.3.4, Insulation Materials VCSNS has discovered that a new 10 CFR 50.48(c)(2)(vii) Approval Request L14 is needed for thermal insulation materials discovered which do not meet the explicit requirements of NFPA 805, Section 3.3.4.

4.3 Precedent 4.3.1 Precedence for Proposed Modification Scope Changes SCE&G inappropriately used performance-based fire risk evaluations to de-scope modifications without obtaining the prior approval from the NRC (Reference 3). This amendment allows the station to modify the transitional license condition found in the operating license. The following plants in addition to others, have received license amendments following the issuance of their NFPA 805 SE for changes to the scope of committed modifications:

Ginna - NFPA 805 Safety Evaluation dated June 25, 2018 (ML18114A025): On June 25, 2018, the NRC issued Amendment No. 127 for the R. E. Ginna Nuclear Power Plant.

The amendment revises the license to delete the modification to install overcurrent protection on its emergency diesel generators which was required as part of Ginna's implementation of its risk-informed, performance-based fire protection program in accordance with paragraph 50.48(c) of Title 10 of the Code of Federal Regulations. In accordance with 10 CFR 50.48(c)(3)(i), the licensee submitted an LAR to revise its fire protection license condition 2.C(3).

Farley - SE dated October 17, 2016 (ML16232A000): On October 17, 2016, the NRC issued Amendment Nos. 205 and 201 for the Joseph M. Farley Nuclear Plant, Units 1 and 2. The amendments update Attachment M, "License Condition Changes";

Attachment S, "Modification and Implementation Items"; and Attachment W, "Fire Probabilistic Risk Analysis Insights," of the previously approved National Fire Protection Association (NFPA) 805 amendment. In accordance with 10 CFR 50.48(c)(3)(i), the licensee submitted a LAR to revise its NFPA 805 transition license conditions.

4.3.2 Precedence for Proposed Changes to the Fundamental Fire Protection Program and Design Elements Reviews of the NFPA 805 implementation identified changes to the license bases that are required to address as-found conditions at the station. The station discovered two

Document Control Desk LAR-16-01490 RC-18-0091 Page 11 of 18 deviations from the NFPA 805 section 3.0, Fundamental Fire Protection Program and Design Elements that were previously approved under 10 CFR 50.48(c)(2)(vii). The following plants in addition to others, have received license amendments for the similar issues:

4.3.2.1 Wiring Above Suspended Ceilings On July 6, 2018, the NRC issued Amendment Nos. 340 and 322 to the D. C. Cook Nuclear Plants (Unit 1 and 2) Operating Licenses. The amendments approve a deviation from the requirements of National Fire Protection Association Standard 805, Section 3.3.5.1, regarding the use of non-plenum listed cables above suspended ceilings, and Section 3.3.5.2, regarding the use of electric metallic tube and embedded/buried polyvinyl chloride conduit.

4.3.2.2 Combustible Insulation H. B. Robinson - SE dated February 3, 2017 (ML16337A264): On February 3, 2017, the NRC issued Amendment No. 249 to the H. B. Robinson Renewed Facility Operating License. The amendment authorized the transition of the fire protection program to a risk-informed, performance-based program based on NFPA 805. Section 3.1.4.5 of the NRC Safety Evaluation discussed H. B. Robinsons use of insulation materials that did not meet NFPA 805 Section 1.6.36 definition of non-combustible materials. NRC evaluated these existing similar insulation materials and concluded it was an acceptable alternative.

McGuire - SE dated September 26, 2013 (ML16077A135): On September 26, 2013, Duke Energy submitted a LAR for the McGuire Nuclear Station Units 1 and 2, to transition the fire protection program to a risk-informed, performance-based program based on NFPA Standard 805. By letter dated September 29, 2016, Duke Energy submitted a response to a request for additional information for similar insulation material concerns and compliance with NFPA 805, Section 3.3.4, Insulation Materials.

On December 6, 2016, the NRC issued Amendment Nos. 291 and 270 to the Renewed Facility Operating License for McGuire Nuclear Station, Units 1 and 2, respectively authorizing the transition of the fire protection program to a risk-informed, performance-based program based on NFPA Standard 805.

4.4 No Significant Hazards Consideration Determination South Carolina Electric & Gas (SCE&G) has evaluated the proposed changes using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

SCE&G proposes a change to the Virgil C. Summer Nuclear Station (VCSNS) Operating License to update the scope of the required NFPA 805 related plant modifications.

SCE&G has included the risk impacts of these changes and has also included risk impacts associated with new items identified since NRC approval of the VCSNS transition to NFPA 805 (Reference NRC Safety Evaluation dated February 11, 2015, ML14287A289).

As required by 10 CFR 50.91(a), SCE&G analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is presented below.

Document Control Desk LAR-16-01490 RC-18-0091 Page 12 of 18

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The purpose of this amendment is to provide updated information associated with the modifications that were described and committed to in the VCSNS License Amendment Request and subsequently approved by the NRC. This amendment also provides updated information related to Nuclear Safety Compliance Strategies (including recovery actions). The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection requirements that are an acceptable alternative to the 10 CFR Part 50, Appendix R, fire protection features (69 FR 33536; June 16, 2004).

Operation of VCSNS in accordance with the proposed amendment does not result in a significant increase in the probability or consequences of accidents previously evaluated. The proposed amendment does not affect accident initiators or precursors as described in the VCSNS Safety Analysis Report (SAR), nor does it adversely alter design assumptions, conditions, or configurations of the facility, and it does not adversely impact the ability of structures, systems, or components (SSCs) to perform their intended function to mitigate the consequences of accidents described and evaluated in the SAR. The proposed amendment does not adversely alter safety-related systems nor affect the way in which safety-related systems perform their functions as required by the accident analysis. The SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition will remain capable of performing the associated design functions.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No Implementation of the new risk-informed, performance-based fire protection licensing basis, with the revised modifications and Nuclear Safety Compliance Strategies complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance contained in RG 1.205, and does not result in new or different kinds of accidents. The requirements in NFPA 805 address only fire protection and the impacts of fire effects on the plant have been evaluated. The proposed amendment does not involve new failure mechanisms or malfunctions that could initiate a new or different kind of accident beyond those already analyzed in the SAR.

Therefore, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed amendment has been evaluated to ensure that risk and safety margins are maintained within acceptable limits. The risk evaluations for plant changes in relation to the potential for reducing a safety margin, were measured

Document Control Desk LAR-16-01490 RC-18-0091 Page 13 of 18 quantitatively for acceptability using the delta risk (i.e., change in core damage frequency and change in large early release frequency) criteria from Section 5.3.5, Acceptance Criteria, of NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-based Fire Protection Program under 10 CFR 50.48(c), as well as the guidance contained in RG 1.205. Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based methods of NFPA 805 do not result in a significant reduction in the margin of safety.

Therefore, this change does not involve a significant reduction in a margin of safety.

4.5 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Document Control Desk LAR-16-01490 RC-18-0091 Page 14 of 18

6.0 REFERENCES

1.

Shawn Williams, NRC, Letter to Thomas Gatlin, Virgil C. Summer Nuclear Station, Unit 1 - Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (TAC No. ME7586), dated February 11, 2015. (ML14287A289)

2.

Public Meeting March 15, 2018 Pre-Application Meeting with South Carolina Electric & Gas Company (SCE&G) to discuss a proposed license amendment regarding NFPA 805 for V. C. Summer, Unit 1, February 23, 2018.

(ML18054A000)

3.

Anne Boland, NRC, Letter to Michael Tschiltz, Recommended Content for License Amendment Requests that Seek Changes to License Conditions That Were Established in Amendments to Adopt National Fire Protection Association Standard 805 But Have Yet to be Fully Implemented, dated March 2, 2016.

(ML16015A416)

4.

NRC letter dated November 22, 2016, Virgil C. Summer Nuclear Station - NRC Problem Identification and Resolution Inspection Report 05000395/2016007 and Notice of Violation (ML16327A378)

5.

SCE&G letter dated November 15, 2011, License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition), (RC-11-0149) (ML14063A455)

6.

SCE&G letter dated October 10, 2012Property "Letter" (as page type) with input value "RC-12-0142, License Amendment Request - LAR-06-00055, License Amendment Request to Adopt NFPA 805, Response to Request for Additional Information" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, (RC-12-0142) (ML12297A218)

7.

SCE&G letter dated August 14, 2014, License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, (RC-14-0129) (ML14227A737)

8.

SCE&G letter dated February 25, 2014, License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, (RC-14-0027) (ML14063A455)

9.

SCE&G letter dated May 2, 2014Property "Letter" (as page type) with input value "RC-14-0067, License Amendment Request - LAR-06-00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., License Amendment Request-LAR-06-00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, (RC-14-0067) (ML14125A274)

10.

SCE&G letter dated December 11, 2014, License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, (RC-14-0196) (ML14350A217)

11.

NRC, Letter dated August 9, 2017, Virgil C. Summer Nuclear Station, Unit 1 -

NRC Integrated Inspection Report 05000395/2017002 and Notice of Violation

12.

SCE&G letter dated September 6, 2017, Virgil C. Summer Nuclear Station (VCSNS), Unit 1 Docket No. 50-395, Operating License No. NPF-12, Reply to Notice of Violation, (RC-17-0126) (ML17249A663)

13.

Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009 (ML090410014)

Document Control Desk LAR-16-01490 RC-18-0091 Page 15 of 18

14.

National Fire Protection Association, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," Standard 805 (NFPA 805), 2001 Edition, Quincy, Massachusetts

15.

U.S. Nuclear Regulatory Commission, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," RG 1.174, Revision 3, January 2018 (ML17317A256)

16.

American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS), "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME/ANS RA-Sa2009, February 2, 2009

17.

U.S. Nuclear Regulatory Commission, "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities, (DELORES-VEWFIRE), Final Report," NUREG-2180, December 2016 (ML16343A058)

18.

U.S. Nuclear Regulatory Commission, "Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database," NUREG-2169 and EPRI 3002002936, January 2015 (ML15016A069)

19.

U.S. Nuclear Regulatory Commission and Electric Power Research Institute, "Refining And Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE)," NUREG-2178/EPRI 3002005578, April 2016 (ML16110A140)

20.

Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Closure of National Fire Protection Association 805 Frequently Asked Question 08-0046: Incipient Fire Detection Systems," November 23, 2009 (ML093220426)

21.

U.S. Nuclear Regulatory Commission, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Regulatory Guide 1.205, Revision 1, December 2009 (ML092730314)

22.

U.S. Nuclear Regulatory Commission, "Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE), Volume 2: Expert Elicitation Exercise for Nuclear Power Plant Fire-Induced Electrical Circuit Failure,"

NUREG/CR-7150, EPRI 3002001989, and BNL-NUREG-98204-2012, May 2014 (ML14141A129)

23.

NRC letter dated December 5, 2016, Virgil C. Summer Nuclear Station - NRC Triennial Fire Protection Inspection Report 05000395/2016010 (ML16344A014)

24.

SCE&G letter dated October 14, 2013, License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, (RC-13-0142) (ML13289A194)

25.

Nuclear Energy Institute, Guidance For Implementing A Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c) NEI 04-02 Revision 2, April 2008 (ML081130188)

Document Control Desk LAR-16-01490 RC-18-0091 Page 16 of 18 7.0 ACRONYMS Acronym Definition AHJ Authority Having Jurisdiction CCDP Conditional Core Damage Probability CDF Core Damage Frequency CLERP Conditional Large Early Release Probability CRIT Factor Criticality Factor CT Current Transformer DG Diesel Generator DID Defense In Depth DROID Deterministic Requirement Open Item Description EEEE Existing Engineering Equivalency Evaluation EOI Emergency Operating Instructions ERFBS Electrical Raceway Fire Barrier System FAQ Frequently Asked Question FHA Fire Hazard Analysis FPEEE Fire Protection Engineering Equivalency Evaluation FPRA Fire Probabilistic Risk Assessment FPP Fire Protection Program FRE Fire Risk Evaluation FSA Fire Safety Analysis FSAR Final Safety Analysis Report HEP Human Error Probability HGL Hot Gas Layer HRA Human Reliability Analysis HRE Higher Risk Evolution HRR Heat Release Rate IGF Ignition Frequency

Document Control Desk LAR-16-01490 RC-18-0091 Page 17 of 18 Acronym Definition ISFSI Independent Spent Fuel Storage Installation KSF Key Safety Function LAR License Amendment Request LERF Large Early Release Frequency LFS Limiting Fire Scenario MCR Main Control Room MOV Motor Operated Valve MSO Multiple Spurious Operation NEI Nuclear Energy Institute NEIL Nuclear Electric Insurance Limited NFPA National Fire Protection Association NRC Nuclear Regulatory Commission NSCA Nuclear Safety Capability Assessment NSP Non-Suppression Probability NSPC Nuclear Safety Performance Criteria NPO Non-Power Operations OMA Operator Manual Action PB Performance-Based PAU Physical Analysis Unit PCS Primary Control Station POS Plant Operational State PRA Probabilistic Risk Assessment PRM Plant Response Model PSA Probabilistic Safety Assessment PWR Pressurized Water Reactor RAI Request for Additional Information RHR Residual Heat Removal

Document Control Desk LAR-16-01490 RC-18-0091 Page 18 of 18 Acronym Definition RI-PB Risk-Informed, Performance-Based RPV Reactor Pressure Vessel SCE&G South Carolina Electric and Gas SER Safety Evaluation Report SF Severity Factor SSA Safe Shutdown Analysis V&V Verification and Validation VFDR Variance From Deterministic Requirement ZOI Zone of Influence 8.0 ATTACHMENTS

1.

Plant Modifications Associated with the NFPA 805 Transition

2.

Updated LAR Table S-1, Plant Modifications Committed

3.

Supplement to LAR Table S-1

4.

Revised RAI Responses

5.

LAR Attachment W - Fire Risk Insights Updates

6.

LAR Attachment C - Fire Area Transition Updates

7.

LAR Attachment G - Recovery Action Transition Updates

8.

LAR Attachment U - Internal Events PRA Quality Supplement

9.

NFPA 805 Chapter 3 Requirements for Approval 10 CFR 50.48(c)(2)(vii)

10.

Operating License Condition Markup

11.

Revised Operating License Condition

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 1 of 13 VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LAR-16-01490 ENCLOSURE 1 ATTACHMENT 1 Plant Modifications Associated with the NFPA 805 Transition

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 2 of 13 1.0

SUMMARY

DESCRIPTION SCE&G requests approval to modify the second Transition License Condition found in VCSNS Operating License (OL) 2.C.(18), Fire Protection, to revise the scope of Engineering Change Request (ECR) 50784 and ECR 50810. This change allows the Variances From Deterministic Requirements (VFDR) impacted by the modification scope change to be resolved using an updated performance-based fire risk evaluation methodology approved by the NRC.

As requested in the NRC guidance provided by letter to NEI (Reference 3), the following Option B information is provided in this Attachment.

i.

A summary of all changes to the modifications ii. A summary of all changes to the PRA models and explanations for each change iii. New, updated versions in their entirety of the:

o License Condition (Attachment M),

o List of plant modifications (Attachment S), and o Summarizing area wide change-in-risk result tables (Attachment W) iv. A statement that the DID and safety margin evaluations associated with the original LAR have been completed on the proposed changes

v. A summary of all accepted PRA methods being used that were not used in the NFPA 805 amendment request and a Reference to the NRC document accepting the method (i.e., the method should have been previously accepted by NRR staff);

vi. A demonstration of the applicability of the accepted method for the configuration and conditions to which it is being applied; vii. A summary of the changes made to the Nuclear Safety Capability Analysis (NSCA) and associated changes to LAR Attachments C and G that reflect any changes in compliance strategies being used on a fire area basis in redline/strikeout; and viii. A justification for the creation of new and/or removal of previously existing VFDRs and Recovery Actions (RAs).

2.0 VCSNS RESPONSE

i.

Summary of all changes to the modifications LAR Table S-1 is presented in Attachment 2. The level of detail in the existing LAR Table S-1 is insufficient to describe the scope changes with a red-line markup of the table. Therefore, contains a Supplement to LAR Table S-1. This Supplement contains a listing of de-scoped modifications for cables that were specified for protection as part of regulatory correspondence but were subsequently eliminated and addressed via the fire risk evaluation process. Cable protection determined unnecessary for deterministic compliance (e.g. by detailed circuit analysis) are not listed in Attachment 3, Supplement to LAR Table S-1. They are included in Enclosure 1, Attachment 6 which contains the update to LAR Attachment C.

As discussed above, the level of detail in the existing LAR Table S-1 is insufficient to describe the scope changes with a red-line markup of the table. Instead Attachment 4 provides the RAI responses (response information provided in italics) describing the affected modifications in detail with provided markups (replaced wording is colored red with strike-through while added

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 3 of 13 wording is colored red and underlined). Attachment 4 contains markups of the following RAIs to reflect the scope changes of the modifications.

Response to FPE RAI 03, transmitted by letter RC-12-0142, dated October 12, 2012 (Reference 6)

Response to PRA RAI 66.03, transmitted by letter RC-14-0129, dated August 12, 2014 (Reference 7) ii.

Summary of all changes to the PRA models and explanations for each change Since the issuance of the NRC Safety Evaluation, the VCSNS Fire PRA has been updated to address the following objectives:

Incorporate into the Fire PRA model the physical plant modifications and procedure changes completed during the NFPA 805 implementation phase of the transition process. This was performed as required by Implementation Item 22 of LAR Attachment S, Table S-2, dated December 11, 2014. The PRA model includes the changes to modification scopes that are being addressed in this LAR, as well as other plant modifications unrelated to the NFPA 805 committed modifications.

Address Updated Industry Guidance:

o Update the credit for the two fire zones at VCSNS with incipient detection (CB06 and CB15) to reflect the latest modeling guidance as specified in NUREG-2180.

o Update generic ignition frequencies and manual non-suppression probabilities with the latest values reported in NUREG-2169.

o Implement updated heat release rate probability distributions and obstructed plume model analyses as described in NUREG-2178.

Resolve issues identified in the model during the update process and introduce model refinements consistent with NRC approved methodologies to improve model realism.

Implementation Phase Changes The Fire PRA model updates include:

Updates in the scenario target sets due to cable re-routes and as-built configurations of the electrical raceway fire barrier system installations originally committed to in the NFPA 805 LAR.

Updates in the ignition frequency calculations due to the addition of new ignition sources to selected fire zones. In addition, the full Fire PRA was updated to reflect the latest generic fire ignition frequencies documented in NUREG-2169 (Reference 18).

Incorporation of the latest Fire PRA technology guidance. Specifically, the VCSNS Safety Evaluation (ML14287A289) was based on the guidance for modeling incipient detection systems documented in Supplement 1 to NUREG/CR-6850, Chapter 13 (based on FAQ 08-0046). This guidance has been superseded by the methods documented in NUREG-2180 (Reference 17). Consequently, the Fire PRA model has been updated to reflect the revised credit for incipient detection following the guidance in NUREG-2180 (Reference 17).

Detailed fire modeling of selected ignition sources that had previously been conservatively modeled. This includes revised heat release rate probability distributions

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 4 of 13 and incorporation of the obstructed plume model in the Fire PRA in selected fire scenarios with electrical cabinets as ignition sources, using the methodologies and heat release rate input data specified in NUREG-2178 (Reference 19). The heat release rate probability distributions and obstructed plume models were implemented primarily to overcome the risk increase associated with the updated credit calculated for the incipient detection system. In addition, the applicable heat release rate probability distributions in NUREG-2178 (Reference 19) and the manual suppression probability curve for the main control room from NUREG-2169 (Reference 18) were incorporated in the main control board fire scenarios evaluated using the guidance provided in NUREG/CR-6850, Appendix L.

Screening of multi compartment scenarios using detailed fire modeling techniques, used elsewhere in the original LAR analysis, for determining hot gas layer temperatures in the exposed or exposing compartments.

Refinement to the internal events model including the Reactor Coolant Pump (RCP) seal Loss of Coolant Accident (LOCA) model based on the RCP seal upgrades (incorporation of the N-9000 RCP seals).

The similarity between these updates and the methods used in the original NFPA 805 LAR Fire PRA supports the classification of these changes as PRA model maintenance activities.

Fire PRA Corrections and Refinements Corrections and refinements in the Fire PRA model are summarized in the following bullets. These changes, as well as the impact of updating the model to address current industry guidance, generated the need for cutset reviews and the implementation of model refinement activities. These changes were also based on methods used in the original NFPA 805 LAR Fire PRA and are therefore considered to be model maintenance type updates.

Correction of the fire initiator fault tree configuration which was preventing the quantification of the risk contribution of fire scenarios associated with the transient initiators. To address this, the fire initiator was inserted into the Fire PRA model logic to allow the affected event sequences to propagate through the model. This correction did not require new accident sequence logic. Additional model reviews and refinements were performed following these updates.

Some of the physical plant modifications committed to during the NFPA 805 process were not properly accounted for in some fire scenarios. Updates were made to reflect the proper credit in the fire scenarios, including verification and updates in protection provided by electrical raceway fire barrier systems and fire-rated cable, as well as model updates and circuit analysis to ensure proper credit in the model for offsite power availability. Data updates were also utilized to provide the most current and realistic fire results. The modifications were credited following the same approaches as those previously used in other fire scenarios.

Additional risk reduction activities included detailed fire modeling, detailed circuit analysis, logic model refinements, expansion of credit for circuit failure mode probabilities using NUREG/CR-7150, Volume 2 for risk significant circuits, and detailed fire human reliability analysis. Each of these updates were performed using methods, techniques and types of data used in other fire scenarios.

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 5 of 13 Refinements in the main control room abandonment model. Originally, the Fire PRA was quantified crediting Main Control Room (MCR) abandonment in all fire scenarios associated with the fire areas designated as alternate shutdown. The updated main control room abandonment model is symptom-based and, for the fire scenarios that lead to a loss of control in the MCR, credits MCR abandonment based on fire impacts on plant safety functions. Specifically, MCR abandonment due to loss of control is specified for fire scenarios which result in the loss of emergency feedwater consistent with the applicable post-fire shutdown procedure.

Refinements in modeling of the multiple spurious operation (MSO) scenarios based upon completion of committed NFPA 805 modifications. Refinements included detailed circuit analysis and logic corrections in the model. Scenario-specific credit for local RCP trip to avoid RCP seal LOCA was introduced into the model as a recovery action. Procedure changes and modeling changes were made to include additional credit for RCP trip from the MCR and locally, as required. These are considered minor refinements to the model.

Fire impacts to some of the containment isolation paths in the model were previously not captured. It should be noted that the criteria for what is considered a containment bypass remained unchanged. However, updates necessary to ensure fire impacts of these valves, including logic refinements with detailed circuit analysis were implemented.

As a result, risk reduction activities requiring model refinements, addition of circuit failure mode probabilities, refinement of human failure events, and procedure changes were necessary. The same approach had been used previously for other containment isolation paths in the model.

Updates to the Human Reliability Analysis (HRA) dependency analysis to comprehensively identify human failure event (HFE) combinations and apply joint human error probabilities (HEPs) to all combinations found using the HRA Calculator with HEPs artificially increased to prevent truncation. This is not a change to the dependency approach, but only the capturing of more or different combinations.

Scenario-specific credit was taken for use of disconnect switches to prevent/mitigate spurious actuations. Procedure changes were implemented to allow credit for use of disconnect switches in select fire areas. This update reflects incorporating procedure changes into the PRA model using the same HRA methods.

Refinements in the modeling of breaker coordination. Additional circuit analysis was performed to explicitly model selected breakers in the Fire PRA. The modeling of these breakers, including the circuit analysis and cable routing necessary to incorporate them as targets in the model was conducted using the same approved methods as had been used in the Fire PRA during the NFPA 805 transition process.

Selected fire scenarios were updated to reflect the presence of non-cable secondary combustibles. These secondary combustibles were incorporated in the Fire PRA by increasing the heat release rate profile previously used in the analysis to reflect the fire intensity of these combustibles. The addition of the secondary combustibles was required for a limited number of fire zones and is primarily associated with combustible insulation and High Density Polyethylene (HDPE) piping.

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 6 of 13 Address Current Industry Approaches The Fire PRA has been updated to incorporate current regulatory methodologies as noted below. These changes incorporate new industry data into the model using the same methodology as the previous analyses and are therefore considered model maintenance items.

NUREG-2180 (Reference 17), Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities, (DELORES-VEWFIRE), December 2016, ML16343A058).

Application of NUREG-2180 is associated with the treatment of incipient detection systems in Fire Areas CB06 and CB15. The update involved a change from the methodology described in NUREG/CR 6850, Supplement 1, Chapter 13, based upon FAQ 08-0046 (Reference 20), to the approach described in NUREG-2180. Specifically, the update consisted of removing the credit for incipient detection calculated using FAQ 08-0046 from the scenarios in Fire Areas CB06 and CB15 where credit was previously applied and replacing it with applicable credit calculated using the guidance in NUREG-2180. The resulting credit for incipient detection varies as a function of the cabinet ventilation and the detection system inside the cabinet. Some cabinets are protected by the incipient detection systems, while other cabinets are protected by the incipient and a spot-type detection system (i.e., a spot type detection system installed inside the cabinet). The update in the treatment of incipient detection resulted in an increase in the risk contribution of the fire scenarios receiving the credit when compared to the risk contribution with credit calculated using FAQ 08-0046. Consequently, additional Fire PRA model refinements were implemented.

NUREG-2169 (Reference 18) and EPRI 3002002936, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database, January 2015 (ML15016A069).

Application of NUREG-2169 is associated with the generic ignition frequencies in the Fire PRA. This update consisted of replacing the generic ignition frequencies documented in Chapter 10 of NUREG/CR-6850 Supplement 1 with the values available in NUREG-2169. This update included all the ignition frequency bins in the Fire PRA and is reflected comprehensively in the risk contribution from all the fire scenarios.

NUREG-2178/EPRI 3002005578 (Reference 19), Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), December 2015 (ML152666A516).

Application of NUREG-2178 is associated with probability distributions for heat release rates in electrical cabinets and the treatment for obstructed fire plume configurations. This update consisted of applying the heat release rate probability distributions documented in NUREG-2178 to the electrical cabinets in Fire Areas CB06 and CB15. Walkdowns were conducted to perform visual examination of the electrical cabinet internals. As part of the walkdowns, the top of each cabinet was inspected for determining the applicability for the obstructed plume model. The updated heat release rates and obstructed plume models were incorporated into this update.

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 7 of 13 Changes to non-Fire PRA Models There are several changes that have been made to the VCSNS PRA models (other than internal fire) which impact the risk results presented in Attachment 5. The most significant change is the total risk estimates for VCSNS now include contributions from internal flooding.

The LERF and Level 2 models have been updated to include the risk contribution from severe accident induced steam generator tube rupture. Various minor enhancements have been made to the PRA models with respect to the modeling of offsite power and the credit obtained from the Alternate Seal Injection System to mitigate certain accident sequences. A full scope peer review was performed in June 2016. As a result, new Facts and Observations were identified.

These are dispositioned with respect to their impact on the Fire PRA in Attachment 8.

In addition, a seismic PRA is in development and is expected to be submitted to the NRC in 2018. The risk results presented in Attachment 5 have sufficient margin to meet RG 1.174 with the expected seismic results.

iii.

New, Updated Versions of License Condition, Plant Modification and Risk Insights LAR Attachments The proposed markup of the existing VCSNS Operating License is presented in Attachment 10, while the revised (clean) copy is provided in Attachment 11.

LAR Table S-1 is presented in Attachment 2. The level of detail in the existing LAR Table S-1 is insufficient to describe the scope changes with a red-line markup of the table. Therefore, contains a Supplement to LAR Table S-1. This Supplement contains a listing of de-scoped modifications for cables that were specified for protection as part of regulatory correspondence but were subsequently eliminated and addressed via the fire risk evaluation process. Cable protection determined unnecessary for deterministic compliance (e.g., by detailed circuit analysis) are not listed in the Supplement to LAR Table S-1 but are included as changes to LAR Attachment C in Attachment 6. contains the updated LAR Attachment W information for total plant risk and Fire Area Risk Summary.

iv.

Defense-In-Depth (DID) and Safety Margin Evaluations The Fire Risk Evaluations were updated for the fire areas impacted by the descoped modifications. A discussion of the changes to each impacted fire area and a summary of the results of the Fire Risk Evaluations are provided below. Additional information on maintenance of defense-in-depth and safety margins for the fire areas impacted by de-scoped modifications is provided in Enclosure 1, Attachment 6.

Fire Area AB01 DROID-AB01.01, 02, 03, 04, 06, 09 Area Cooling for Charging Pump A and B Rooms Cables required for operation of area cooling for the Charging Pump A & B rooms are protected by one-hour rated Electric Raceway Fire Barrier Systems (ERFBS). Full automatic suppression is required for the one-hour rated fire barrier to achieve deterministic compliance. The change in modification scope (no longer expanding the

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 8 of 13 automatic suppression to the entire fire zone) requires this DROID to be resolved with a Performance-Based Fire Risk Evaluation, since ECR 50810 did not modify the plant to achieve full automatic suppression coverage. Automatic fire suppression is no longer credited in the Nuclear Safety Capability Assessment (NSCA) for Fire Zone AB01.09.

DROID-AB01.01, 02, 03, 04, 06, 09 Charging Pumps A & B Cables required for operation of Charging Pumps A & B are protected by one-hour rated Electric Raceway Fire Barrier Systems. Full automatic suppression is required for the one-hour rated fire barrier to achieve deterministic compliance. The change in modification scope (no longer expanding the automatic suppression to the entire fire zone) requires this DROID to be resolved with a performance-based fire risk evaluation, since ECR 50810 did not modify the plant to achieve full automatic suppression coverage. Automatic fire suppression is no longer credited in the NSCA for Fire Zone AB01.09.

DROID-AB01.01, 02, 03, 04, 06, 09 Charging Pumps A & B Mini-Flow Isolation ECR 50784 modified circuit CS C 82B to replace portions of the circuit with a three-hour fire rated cable. The new fire rated cable terminates at splice boxes located in Fire Zones AB01.04 and AB01.09. The original scope of ECR 50784 prescribed the installation of an Electric Raceway Fire Barrier System to protect the splice box and the remainder of the non-fire rated circuit in Fire Zones AB01.04 and AB01.09. The ECR scope has been modified to remove the scope of work to protect the splice boxes and non-fire rated circuits. This change in modification scope requires this DROID to be resolved with a performance-based fire risk evaluation, since ECR 50784 did not fully result in complete three-hour fire rated barrier protection for cables associated with the Charging Pump Mini-Flow Isolation Valves. The portion of the circuit that is protected with a one-hour electric raceway fire barrier system is not deterministically compliant, since the fire zone does not contain full automatic suppression coverage.

AB01.21 - No DROID Impacted Both the Fire PRA and NSCA models do not credit automatic fire suppression in Fire Zones AB01.21.01 and AB01.21.02. There is no DROID associated with automatic fire suppression in Fire Zones AB01.21.01 and AB01.21.02. Therefore, the change in scope to ECR 50810 to not include enhancements to the automatic suppression system in AB01.21.01 and AB01.21.02 has no impact on the NFPA 805 program, since the system is not needed to support deterministic compliance, for risk reduction or to maintain sufficient defense-in-depth.

Fire Risk Evaluation Conclusions AB01 - Change in Risk The fire risk evaluation concluded that the change in risk is acceptable and that defense-in-depth and safety margins are maintained without the modifications. The new as-built risk results are provided in Attachment 5.

AB01 - Defense-in-Depth Evaluation The defense-in-depth evaluation for Fire Area AB01 has been updated in the performance-based fire risk analysis to reflect the as-built configuration of VCSNS,

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 9 of 13 which includes the adjustments in modification scope for Fire Area AB01. The performance-based fire risk analysis uses a qualitative analysis of the defense-in-depth echelon balance, along with insights from the Fire PRA to ensure adequate defense-in-depth is maintained. The defense-in-depth evaluation for Fire Area AB01 in the performance-based fire risk analysis concludes that adequate defense-in-depth is maintained.

AB01 - Safety Margin Evaluation The safety margin evaluation for Fire Area AB01 has been updated in the Performance-Based Fire Risk Analysis to reflect the as-built configuration of VCSNS, which includes the adjustments in modification scope for Fire Area AB01. The change in modification scope does not affect the safety margin evaluation for the fire area. The fire scenarios were reviewed to verify that the feed and bleed method is not the sole method of cooling the reactor coolant system as required by 10 CFR 50.48(c)(iii).

Fire Area CB01 DROID-CB01 Source Range Flux Monitors The NSCA model requires at least one source range flux detector to be available for deterministic compliance, while the Fire PRA model does not credit these components for success. The original scope consisted of protecting the associated conduits and junction boxes for cable NI A 198D with an Electrical Raceway Fire Barrier System (ERFBS) to protect INI0032. However, the final scope of ECR 50784 did not include a ERFBS for cable NI A 198D. This DROID is now resolved with a performance-based fire risk evaluation per NFPA 805 4.2.4.2 Use of Fire Risk Evaluation.

DROID-CB01 Spurious Containment Isolation Phase A Signal The NSCA report identifies a concern where fire damage could simulate a manual initiation of a Containment Isolation Phase A Signal. The Fire PRA model also considers this failure. The original scope of ECR 50784 prescribed ERFBS to be placed on the cable trays which contained cable SG D 14B. However, the final scope of ECR 50784 did not include a ERFBS for cable SG D 14B. This DROID is now resolved with a performance-based fire risk evaluation per NFPA 805 4.2.4.2 Use of Fire Risk Evaluation.

Fire Risk Evaluation Conclusions CB01 - Change in Risk The fire risk evaluation concluded that the change in risk is acceptable and that defense-in-depth and safety margins are maintained without the modifications. The new as-built risk results are provided in Attachment 5.

CB01 - Defense-in-Depth Evaluation The defense-in-depth evaluation for Fire Area CB01 has been updated in the Performance-Based Fire Risk Analysis to reflect the as-built configuration of VCSNS, which includes the adjustments in modification scope for Fire Area CB01. The Performance-Based Fire Risk Analysis uses a qualitative analysis of the defense-in-depth echelon balance, along with insights from the Fire PRA to ensure adequate

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 10 of 13 defense-in-depth is maintained. The defense-in-depth evaluation for Fire Area CB01 in the Performance-Based Fire Risk Analysis concludes that adequate defense-in-depth is maintained.

CB01 - Safety Margin Evaluation The safety margin evaluation for Fire Area CB01 has been updated in the Performance-Based Fire Risk Analysis to reflect the as-built configuration of VCSNS, which includes the adjustments in modification scope for Fire Area CB01. The change in modification scope does not affect the safety margin evaluation for the fire area. The fire scenarios were reviewed to verify that the feed and bleed method is not the sole method of cooling the reactor coolant system as required by 10 CFR 50.48(c)(iii).

Fire Area IB17 DROID-IB17 XSW1DA Failure The NSCA and Fire PRA models are impacted by cables ES M 83X and ES M 94X routed through Fire Area IB17 which, if fire damaged, could cause a failure to obtain power from 7.2kV Switchgear XSW1DA. These cables were not modified by ECR 50784 or ECR 50800, so the DROID is now resolved with a performance-based fire risk evaluation per NFPA 805 4.2.4.2 Use of Fire Risk Evaluation.

Fire Risk Evaluation Conclusions IB17 - Change in Risk The fire risk evaluation concluded that the change in risk is acceptable and that defense-in-depth and safety margins are maintained without the modifications. The new as-built risk results are provided in Attachment 5.

IB17 - Defense-in-Depth Evaluation The defense-in-depth evaluation for Fire Area IB17 has been updated in the Performance-Based Fire Risk Analysis to reflect the as-built configuration of VCSNS, which includes the adjustments in modification scope for Fire Area IB17. The Performance-Based Fire Risk Analysis uses a qualitative analysis of the defense-in-depth echelon balance, along with insights from the Fire PRA to ensure adequate defense-in-depth is maintained. The defense-in-depth evaluation for Fire Area IB17 in the Performance-Based Fire Risk Analysis concludes that adequate defense-in-depth is maintained.

IB17 - Safety Margin Evaluation The safety margin evaluation for Fire Area IB17 has been updated in the Performance-Based Fire Risk Analysis to reflect the as-built configuration of VCSNS, which includes the adjustments in modification scope for Fire Area IB17. The change in modification scope does not affect the safety margin evaluation for the fire area. The fire scenarios were reviewed to verify that the feed and bleed method is not the sole method of cooling the reactor coolant system as required by 10 CFR 50.48(c)(iii).

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 11 of 13 Fire Area IB25 DROID-IB25.04 Loss of Charging Pump Suction from VCT Two cables associated with valve LCV0115E are routed in IB25.04 which impact both the NSCA and the Fire PRA models. Cable CS C 302B was partially replaced with a three-hour fire rated cable by the same modification. However, the splice box where the new fire rated cable for CS C 302B terminates in IB25.04 was not protected as originally prescribed by ECR 50784. Thus, the DROID is now resolved with a performance-based fire risk evaluation per NFPA 805 4.2.4.2 Use of Fire Risk Evaluation.

DROID-IB25.04 Loss of Charging Pump Minimum Flow Path The NSCA and Fire PRA models are both impacted by a potential fire-induced spurious closure of the Charging Pump Mini-Flow Valves in Fire Zone IB25.04. The DROID identifies fire damage potentially causing spurious closures of the A and C Charging Pump Mini-Flow Valves. ECR 50784 prescribed for cable CS C 72B to be partially replaced with a three-hour fire rated cable, with the terminating junction boxes and remaining cable in the affected fire zones protected with ERFBS. However, the junction box and remaining cable in Fire Zone IB25.04 were not protected. Therefore, the DROID is now resolved with a performance-based fire risk evaluation per NFPA 805 4.2.4.2 Use of Fire Risk Evaluation.

DROID-IB25.06, 07 XSW1DB Failure The NSCA and Fire PRA models are both impacted by a possible fire-induced spurious closure of the alternate offsite power feed to XSW1DB. Spurious operation of the breaker could cause a loss of XSW1DB. ECR 50784 prescribed for cable ES M 73X to be partially replaced with a three-hour fire rated cable, with the terminating junction boxes and remaining cable in the affected fire zones protected with ERFBS. However, the junction box and remaining cable in Fire Zone IB25.06 were not protected.

Therefore, the DROID is now resolved with a performance-based fire risk evaluation per NFPA 805 4.2.4.2 Use of Fire Risk Evaluation.

DROID-IB25.06, 07 XSW1DB Failure The NSCA and Fire PRA models are both impacted by a possible fire-induced failure to open the normal offsite power feed to XSW1DB. A failure to open the breaker could prevent other sources of power from energizing XSW1DB. ECR 50784 prescribed for cable ES M 63X to be partially replaced with a three-hour fire rated cable, with the terminating junction boxes and remaining cable in the affected fire zones protected with ERFBS. However, the junction box and remaining cable in Fire Zone IB25.06 were not protected. Therefore, the DROID is now resolved with a performance-based fire risk evaluation per NFPA 805 4.2.4.2 Use of Fire Risk Evaluation.

Fire Risk Evaluation Conclusions IB25 - Change in Risk The fire risk evaluation concluded that the change in risk is acceptable and that defense-in-depth and safety margins are maintained without the modifications. The new as-built risk results are provided in Attachment 5.

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 12 of 13 IB25 - Defense-in-Depth Evaluation The defense-in-depth evaluation for Fire Area IB25 has been updated in the Performance-Based Fire Risk Analysis to reflect the as-built configuration of VCSNS, which includes the adjustments in modification scope for Fire Area IB25. The Performance-Based Fire Risk Analysis uses a qualitative analysis of the defense-in-depth echelon balance, along with insights from the Fire PRA to ensure adequate defense-in-depth is maintained. The defense-in-depth evaluation for Fire Area IB25 in the Performance-Based Fire Risk Analysis concludes that adequate defense-in-depth is maintained.

IB25 - Safety Margin Evaluation The safety margin evaluation for Fire Area IB25 has been updated in the Performance-Based Fire Risk Analysis to reflect the as-built configuration of VCSNS, which includes the adjustments in modification scope for Fire Area IB25. The change in modification scope does not affect the safety margin evaluation for the fire area. The fire scenarios were reviewed to verify that the feed and bleed method is not the sole method of cooling the reactor coolant system as required by 10 CFR 50.48(c)(iii).

v.

A summary of all accepted PRA methods being used that werent used in the NFPA 805 amendment request and a Reference to the NRC document accepting the method (i.e., the method should have been previously accepted by NRR staff)

The following accepted methods are used in the Fire PRA:

o Updates for incipient detection system in Fire Areas CB06 and CB15 to reflect the latest modeling guidance as specified in NUREG-2180 (Reference 17).

o Updates to the generic ignition frequencies and manual non-suppression probabilities with the latest values reported in NUREG-2169 (Reference 18).

o Updates to reflect the heat release rate probability distributions and obstructed plume model analyses as described in NUREG-2178 (Reference 19).

These guidance documents were not approved for use during the original development and review of the VCSNS Fire PRA. The use of the latest industry guidance, either in the base model or as an implementation item, is reflected in more recent NFPA 805 Safety Evaluations, including:

o Ginna - NFPA 805 Safety Evaluation dated June 25, 2018 (ML18114A025).

o Farley - SE dated October 17, 2016 (ML16232A000).

vi.

A demonstration of the applicability of the accepted method for the configuration and conditions to which it is being applied The new methods applied, as noted in item v above are associated with refinement of data input via NRC issued NUREGs. The new panel heat release rates and obstructed plume methodology (NUREG-2178) along with the updated ignition frequencies and manual non-suppression probabilities (NUREG-2169) and updated incipient detection credit (NUREG-2180) are applicable as specified by the NRC in the issuance of the associated NUREGs.

Document Control Desk - Attachment 1 LAR-16-01490 RC-18-0091 Page 13 of 13 vii.

A summary of the changes made to the Nuclear Safety Capability Analysis (NSCA) and associated changes to LAR Attachments C and G that reflect any changes in compliance strategies being used on a fire area basis in redline/strikeout Redline/Strikeout changes to LAR Attachment C, B-3 Table Fire Area Transition is included in. The revisions reflect the following changes to the B-3 Table:

De-scoped modifications Changes in Compliance Strategy (e.g, areas that changed from Compliance with Section 4.2.4.1 Use of Fire Modeling of NFPA 805 to 4.2.4 Performance-Based Approach of NFPA 805)

New or revised DROIDS (VFDRs)

Removal of DROIDS (VFDRs) due to completion of modifications Redline/Strikeout changes to LAR Attachment G Recovery Action Transition is included in, Attachment 7.

viii.

A justification for the creation of new and/or removal of previously existing Variances from Deterministic Requirements (VFDRs) and Recovery Actions (RAs), Attachment 6 and 7 provide the justifications for the changes to VFDRs and recovery actions.

Document Control Desk - Attachment 2 LAR-16-01490 RC-18-0091 Page 1 of 2 VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LAR-16-01490 ENCLOSURE 1 ATTACHMENT 2 Updated LAR Table S-1, Plant Modifications Committed

Document Control Desk - Attachment 2 LAR-16-01490 RC-18-0091 Page 2 of 2 Table S-1 Plant Modifications Committed Item Rank Location Problem Statement Proposed Change In Fire PRA Comp Measure Risk Informed Characterization Completion ECR 50784:

NFPA 805 Circuit/Tubing Protection Low As Defined Additional insights gained during performance of NFPA 805 analysis defining circuit and equipment interactions.

Provide protection of tubing/circuits from the effects of fire. Note 1 Yes Yes Protection in the form of 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire rating is being provided for select circuits identified through the NSCA.

2015 ECR 50799:

NFPA 805 RCP Seal Replacement Medium RB412 Improvement in station equipment to address Loss of Seal Cooling/LOCA scenarios for RCP Seals.

Provide lower leakage RCP Seals [Outage].

Yes None Alternate Seal Injection obviates much of the benefit of this modification. This would be ranked High if not for Alternate Seal Injection.

2015 ECR 50810:

NFPA 805 Hazard Protection High As Defined Fire protection feature enhancements.

Provide mitigation strategies to address fire initiators or limit fire propagation. Note 1 Yes Yes A sensitivity study for the fire PRA showed that this modification was highly important.

2015 ECR71588:

NFPA 805 Penetration Seal Documentation Low Various Improve documentation of penetration seal designs to penetration tests.

Document updates to include improved penetration details and alignment with vendor tests.

Yes None Integrity of fire barriers is maintained by the quality of penetration seal installations vs. fire test configurations (important to fire scenario development).

2015 ECR 50856:

NFPA 805 Communication Medium As Defined Improve availability and reliability of station communication system(s) during fire scenarios.

Provide alternate backup, protected communication system to support fire event.

No None Communication is implicitly considered in credit for Fire PRA operator actions.

However, many are performed in the control room where communication is not threatened by fire.

2015 Note: ECR71588 is not a plant modification. This ECR was added to Table S-1 to emphasize the importance and size of the scope.

Note 1The de-scope portions of ECRs 50784 and 50810 is shown in the Supplement to LAR Table S-1 table found in Attachment 3 of RC-18-0091.

Document Control Desk - Attachment 3 LAR-16-01490 RC-18-0091 Page 1 of 2 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LAR-16-01490 ENCLOSURE 1 ATTACHMENT 3 Supplement to LAR Table S-1

Document Control Desk - Attachment 3 LAR-16-01490 RC-18-0091 Page 2 of 2 Supplement to Table S-1 Fire Area Fire Area Description ECR Withdrawn Scope De-Scoped Original Cable IDs New Cable ID For Affected DROID(s)

Affected DROID(s)

AB01 AB General Area, All Elevations (ex WPAA) 50784 Protect Circuits for XVT08109A and XVT08109B CS C 72B CS C 82B CS C 76B CS C 85B CS C 86B DROID-AB01.01, 02, 03, 04, 06, 09-10 50810 AB 400' el. Sprinkler System - Extend Coverage N/A DROID-AB01.01, 02, 03, 04, 06, 09-05 DROID-AB01.01, 02, 03, 04, 06, 09-06 DROID-AB01.01, 02, 03, 04, 06, 09-10

`

50810 AB 463' el. Sprinkler System - Extend Coverage N/A None CB01 CB General Area 412, 425 West 50784 Protect Circuits for INI00032 NI A 198D DROID-CB01-01 50784 Protect Circuits for XCP06104-CS-SG02A and -SG02B SG D 14B DROID-CB01-32 IB17 IB ESF SWGR Cooling Unit Room B 50784 50800 Modify cables associated with XSW1DA-U01 and -U015 ES M 83X, ES M 94X DROID-IB17-04 IB25 IB General Area 412, 436/ WPAA 463 50784 Protect Circuits for LCV00115E CS C 302B CS C 351B CS C 352B DROID-IB25.04-04 50784 Protect Circuits for XVT08109A:Open:Open CS C 72B CS C 75B CS C 76B DROID-IB25.04-10 50784 Protect Circuits for XSW1DB-U01 ES M 73X ES M 465X DROID-IB25.06, 07-09 50784 Protect Circuits for XSW1DB-U16 ES M 63X ES M 461X DROID-IB25.06, 07-10

Document Control Desk - Attachment 4 LAR-16-01490 RC-18-0091 Page 1 of 8 VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LAR-16-01490 ENCLOSURE 1 ATTACHMENT 4 Revised RAI Responses

Document Control Desk - Attachment 4 LAR-16-01490 RC-18-0091 Page 2 of 8 The level of detail in the existing LAR Table S-1 is insufficient to describe the scope changes with a red-line markup of the table. Instead, the RAI responses (response information provided in italics) describing the affected modifications in detail are provided with markups (replaced wording is colored red with strike-through while added wording is colored red and underlined).

FPE RAI 03 Response from RC-12-0142, dated October 12, 2012 (Reference 6)

LAR Attachment S Plant Modifications Committed ECR 50810 is supposed to provide mitigation strategies to address fire initiators or limit fire propagation.

a.

Provide more detail for this modification(s).

b.

Describe the modification, including location.

c.

Describe the applicable design standards.

d.

Describe the elements of defense-in-depth (DID) that are being satisfied.

e.

This modification is identified as being in the Fire PRA. Describe specifically what is in the Fire PRA.

SCE&G Response The scope of the ECR includes Fire Protection feature enhancements identified while transitioning the stations Fire Protection Program to one based on NFPA 805. The scope of the modification package presently includes, as applicable:

Transient Materials: Signage/ Floor Demarcations (Admin Controls)

• AB 400 Sprinkler System Rework NFPA 13 AB 463 Sprinkler System Rework (Minor) NFPA 13 Junction Box/ Pull Box Pillows Upgrades ** (Embedded Conduit)

Flammable Liquid Storage Cabinets (Seismic Considerations)

Portable Fire Extinguisher Additions NFPA 10 Smoke Detectors (Based on final walkdowns of engineering documentation) NFPA 72E The elements of defense in depth being satisfied range from fire prevention to fire suppression and detection.

Concerning the Statement: "In the Fire PRA" relates to crediting fire protection features that meet specified standards, which would include NFPA Codes. In selected cases, enhancements to existing systems and features were identified during this review and are being addressed accordingly.

• Signage/Floor Demarcations were established per ECR 50810 but were later determined to not be required. The signage/floor demarcations were determined to be unnecessary as the compliance basis for the affected fire areas was changed from Section 4.2.4.1 Use of Fire Modeling to Section 4.2.4 Performance-Based Approach of NFPA 805.

• • Fire Protection Engineering Equivalency Evaluation TR0780E-006 Attachment FEAT 02 provided for two approaches for protecting junction boxes / pull boxes. One method is to use mineral wool blankets. Alternatively, some junction boxes / pull boxes were protected using CS-195+ board attached to the underside of the junction boxes / pull boxes covers. ECR 50810 used both methods. This item is revised for clarification and is not a descoped modification.

Document Control Desk - Attachment 4 LAR-16-01490 RC-18-0091 Page 3 of 8 PRA RAI 66.03 Response from RC-14-0129, dated August 12, 2014 (Reference 7)

In a letter dated May 2, 2014Property "Letter" (as page type) with input value "RC-14-0067, License Amendment Request - LAR-06-00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., (ADAMS Accession No. ML14125A274) the licensee responded to PRA RAI 66.02 and stated that "additional cable protection modifications were added to protect the CREP panels in CB04, CB06, CB15, and CB17" and "additional cable protection modifications were added to the latest fire PRA model to protect several additional functions including prevention of spurious ESFAS, prevention of RWST draindown, isolation of spray suction, and EDG loading." It does not appear, however, that these modifications are included in the updated LAR Attachment C provided in the letter dated February 25, 2014 (ADAMS Accession No. ML14063A455), nor are they described in the updated LAR Attachment S, Table S-1, provided in the letter dated October 14, 2013 (ADAMS Accession No. ML13289A194). Identify which of the ECR modification(s) in LAR Attachment S, Table S-1 includes these "additional cable protection modifications." In addition, describe all the individual modifications included under this ECR modification(s), and describe where each of these individual modifications is specifically identified in the LAR to either make a VFDR (DROID) deterministically compliant or to only reduce risk.

SCE&G Response The "additional cable protection items" discussed in SCE&G's response to PRA RAI 66.02 (RC-14-0067) did not result in new Engineering Change Request (ECR) modifications being added to LAR Attachment S, Table S-1. Instead, the scope of two existing ECR modifications have been modified as necessary to include these items.

The two ECRs that contain the scope of these modifications are:

ECR 50810, NFPA 805 Hazard Protection ECR 50784, NFPA 805 Circuit/Tubing Protection Additional information regarding the scope of each of these modifications is provided below including identification of why they were required, either for a DROID (VFDR), or for risk reduction.

Additional Cable Protection Items discussed in PRA RAI 66.02 Response that are included in the scope of ECR 50810:

The response to RAI 66.02 states that cable protection was applied to equipment controls from the Control Room Evacuation Panel (CREP). This was to address potential loss of CREP controls for fires in CB04, CB06, CB15, and CB17. The CREP panel protection cables are in the FRANX database and documented in the Quantification notebook in the table "Firelmpact Data - Mods". The scope of this modification has been added to ECR 50810.

It should be noted that although circuits associated with this modification are described as "protected" in the PRA RAI 66.02 response, this is only FPRA modeling terminology.

ECR 50810 will independently power the EFW isolation valves (IFV-3531, 3541, 3551, 3536, 3546 and 3556) from their respective CREP panels (XPN-7006A1B) and isolate power from the Control Room. This is to ensure Emergency Feedwater control from the CREP panels is available. The goal is preservation of equipment function from the CREP. The function of the circuits listed in Table 1 will be protected, not the cables themselves.

Document Control Desk - Attachment 4 LAR-16-01490 RC-18-0091 Page 4 of 8 This modification was identified to reduce risk in the FPRA. ECR 50810 is in the design phase and is scheduled to be completed by the end of 2015.

Table 1 - FPRA RAI 66.03 Response Fire Area DROID ID Cable Plant ECR CB04, CB06, CB15 N/A (Risk Reduction)

EF W 21B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 22B 50810 CB15 N/A (Risk Reduction)

EF W 23B 50810 CB15 N/A (Risk Reduction)

EF W 24B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 56B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 57B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 61A 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 62A 50810 CB15 N/A (Risk Reduction)

EF W 63A 50810 CB15 N/A (Risk Reduction)

EF W 64A 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 86A 50810 CB15 N/A (Risk Reduction)

EF W 89A 50810 CB15 N/A (Risk Reduction)

EF W 91B 50810 CB15 N/A (Risk Reduction)

EF W 92B 50810 CB15 N/A (Risk Reduction)

EF W 93B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 94B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 95B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 97B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 98B 50810 CB15 N/A (Risk Reduction)

EF W 103B 50810 CB15 N/A (Risk Reduction)

EF W 104B 50810 CB15 N/A (Risk Reduction)

EF W 109B 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 110B 50810 CB15 N/A (Risk Reduction)

EF W 111A 50810 CB15 N/A (Risk Reduction)

EF W 112A 50810 CB15 N/A (Risk Reduction)

EF W 113A 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 114A 50810 CB15 N/A (Risk Reduction)

EF W 115A 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 116A 50810 CB15 N/A (Risk Reduction)

EF W 117B 50810 CB15 N/A (Risk Reduction)

EF W 122A 50810 CB15 N/A (Risk Reduction)

EF W 123A 50810 CB04, CB06, CB15 N/A (Risk Reduction)

EF W 128A 50810 CB15 N/A (Risk Reduction)

EF W 133B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7003A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7005A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7012A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7014A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7017A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7018A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7019A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7020A 50810

Document Control Desk - Attachment 4 LAR-16-01490 RC-18-0091 Page 5 of 8 Fire Area DROID ID Cable Plant ECR CB15, CB17.01 N/A (Risk Reduction)

EF W7021A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7022A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7023A 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7024B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7025B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7026B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7027B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7028B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7029B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7030B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7031B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7032B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7034B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7035B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7036B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7037B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7039B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7041B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7042B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7043B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7044B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7045B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7046B 50810 CB15, CB17.01 N/A (Risk Reduction)

EF W7047B 50810 CB15, CB17.01 N/A (Risk Reduction)

MC E7004A 50810 CB15, CB17.01 N/A (Risk Reduction)

MS W7019A 50810 CB04, CB06 N/A (Risk Reduction)

SG J 31A 50810 CB04, CB06 N/A (Risk Reduction)

SG J 32B 50810 Additional Cable Protection Items discussed in FPRA RAI 66.02 Response that are included in the scope of ECR 50784:

SCE&G's PRA RAI 66.02 response discussed cable protection of additional items to preserve several nuclear safety functions. It should be noted that although these items were described as "additional cable protection items, "some of them were only newly credited in the FPRA and have been included in the scope of ECR 50784 to support the Nuclear Safety Capability Assessment (NSCA) since the original LAR submittal. Table 2 includes the requested details for these modifications. ECR 50784 is in the implementation phase and all work is scheduled to be completed by the end of 2015.

Table 2 - FPRA RAI 66.03 Response Fire Area DROID ID Cable Notes Plant ECR AB01.04 DROID-AB01.01, 02, 03, 04, 06, 09-10 CS C 82B 50784 AB01.06 DROID-AB01.01, 02, 03, 04, 06, 09-10 CS C 72B 50784

Document Control Desk - Attachment 4 LAR-16-01490 RC-18-0091 Page 6 of 8 Fire Area DROID ID Cable Notes Plant ECR AB01.09 DROID-AB01.01, 02, 03, 04, 06, 09-10 CS C 82B 50784 CB02 DROID-CB02-10 ED E 34B Protection of ED E 34B preserves desired function. EM C 287B does not need protection as identified in the RAI 66.02 response.

50784 CB05 DROID-CB05-04 ED E 23A ED E 24A Protection of ED E 23A and ED E 24A preserves the desired equipment function. EM C 281A and EM C 282A do not need protection as identified in the RAI 66.02 response.

50784 IB25.04 DROID-IB25.04-08 SI C 212B SP C 92B 50784 IB25.04 DROID-IB25.04-04 CS C 302B CS C 305B 50784 IB25.04 DROID-IB25.04-10 CS C 72B 50784 IB25.06.02 DROID-IB25.06, 07-09 ES M 73X 50784 IB25.06.02 DROID-IB25.06, 07-10 ES M 63X 50784 IB25.06.02 N/A (For Risk Reduction)

RC M 15X RC M 25X RC M 35X 50784 Full Scope of ECR 50784 ECR 50784 consists of circuit protection modifications including replacement of cable with fire rated cable, and installation of fire wrap. To address the request for all modifications included in ECR 50784, Table 3 contains the remaining circuit protection modifications that were not identified in Table 2. The scope of ECR 50784 agrees with the B-3 Table submitted under RC-14-0027 except for DROID-IB25.04.02. It was determined during ECR 50784 implementation that circuit protection is not required because redundant indication is available.

Table 3 - FPRA RAI 66.03 Response Fire Area DROID ID Cable Plant ECR AB01.10 DROID-AB.10, 13, 14, 15, 16, 17-05 RC E 5XB 50784 AB01.08.02 DROID-AB01.08-03 CS C103A 50784 AB01.10 DROID-AB.10, 13, 14, 15, 16, 17-01 MS U 101D SI U 3D 50784 AB01.18.01 DROID-AB01.18, 19-03 SI U 3D 50784 CB01.01 DROID-CB01-01 NI A 198D 50784 CB01.01 DROID-CB01-18 CS C 52B CS C 62B 50784

Document Control Desk - Attachment 4 LAR-16-01490 RC-18-0091 Page 7 of 8 Fire Area DROID ID Cable Plant ECR CB01.01 DROID-CB01-32 SG D 14B 50784 CB02 DROID-CB02-07 CS C 52B CS C 62B 50784 CB02 DROID-CB02-42 EF U 55A 50784 CB12 DROID-CB12-19 CS C 42B CS C 52B CS C 62B 50784 CB20 DROID-CB20-04 CS C 42B 50784 IB25.01.03 DROID-IB25.01.03,.04,.05-01 EF U 4A EF U 55A 50784 IB25.01.03 DROID-IB25.01.03,.04,.05-02 MS W 251A MS W 256A 50784 IB25.06-02 DROID-IB25.06, 07-05 VL C 12C 50784 AB01.21.02 N/A (Risk Reduction Only)

CS W 304XD 50784 CB10 N/A (Risk Reduction Only)

ES M 94X 50784 CB15 N/A (Risk Reduction Only)

ES E 108X 50784 Additionally, the NSCA credited ECR 50800 in SCE&G Letter RC-14-0027 (Reference 8) to resolve DROID-IB17-04. ECR 50800 and ECR 50784 did not fully address circuit protection of cables ES M 83X and ES M 94X in Fire Area IB17, so DROID-IB17-04 is resolved using the Performance-Based Fire Risk Evaluation methodology approved in VCSNS NFPA 805 SE (Reference 1) pending NRC approval of this LAR.

Full Scope of ECR 50810 The purpose of plant modification ECR 50810 is to provide plant enhancements and address items identified during the NFPA 805 transition process. In addition to the CREP panel protection, the current scope includes replacing sections of existing sprinkler systems, installation of additional portable fire extinguishers, raising Post Indicator Valves, installation of lock boxes and storage boxes, anchoring flammable liquid storage cabinets and modifications to address LERs 2011-001, 2011-002 and 2013-005.

The original fire suppression (sprinkler) modification scope prescribed additional sprinkler coverage in the Auxiliary Building 400 and 463 elevations. The scope was subsequently removed in a revision to the ECR. This modification was to provide Fire Zone AB01.09 with full suppression and was included as part of the justification for NFPA 805 Fire Protection Engineering Equivalency Evaluation FPEEE-AB01-01, Auxiliary Building - Lack of 20 ft separation and automatic suppression (4.2.3.3(b) criteria) found in Technical Report TR0780E-001, Attachment AB01-01. Since the modification was not performed, Attachment AB01-01 of the Fire Protection Engineering Equivalency Evaluation has been withdrawn and automatic suppression and detection are no longer credited in Fire Zone AB01.09 for the NSCA. Detection remains credited in Fire Zone AB01.09 in the Fire PRA model. The following DROIDS for AB01.01, 02, 03, 04, 06, 09 appearing in the Attachment C NEI 04-02 Table B-3 found in SCE&G Letter RC-14-0027 (Reference 8) are now resolved using the Performance-Based Fire Risk Evaluation methodology approved in VCSNS NFPA 805 SE (Reference 1) pending NRC approval of this LAR:

Document Control Desk - Attachment 4 LAR-16-01490 RC-18-0091 Page 8 of 8 DROID-AB01.01, 02, 03, 04, 06, 09-05 DROID-AB01.01, 02, 03, 04, 06, 09-06 DROID-AB01.01, 02, 03, 04, 06, 09-10 The modification scope for the Auxiliary Building 463 elevation, in Fire Zones AB01.21.01 and AB01.21.02, was intended to address perceived deficiencies with the existing suppression system. The existing pre-action sprinkler system provides coverage for a portion of these fire zones, as shown on Drawing E-023-011. However, the automatic suppression systems are not credited in both the NSCA and the Fire PRA models for Fire Zones AB01.21.01 and AB01.21.02. Therefore, no improvements to the suppression system are necessary. No DROIDS were affected by this descoping.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 1 of 28 VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LAR-16-01490 ENCLOSURE 1 ATTACHMENT 8 LAR Attachment U - Internal Events PRA Quality Supplement

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 2 of 28 The following table provides a review of the finding level Internal Events Facts and Observations (F&Os) to the June 2016 Peer Review. The Peer Review was based on ASME/ANS Ra-Sb-2009 for internal events and ASME/ANS Ra-Sb-2013 for internal flooding as well as RG 1.200, Rev. 2 (Reference 13). The review provides an assessment of the impact on the Fire PRA model and insights for the Internal Events PRA Model. The F&Os associated with Supporting Requirements Met at Category II and Category III are noted as such.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 3 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 1-01 MU-B3 Finding There is no clear statement in PSA-08 that PRA updates and upgrades should be performed in accordance with the PRA Standard (This F&O originated from SR MU-B3)

PRA changes may be performed in a manner that is not consistent with the PRA Standard.

Recommend stating specifically that PRA updates and upgrades be performed in accordance with the PRA Standard.

This SR is met at CC-I-III.

Procedure change with no fire PRA impact.

Procedure has been updated to state updates and upgrades should be performed to the ASME standards. The issue is resolved.

1-02 MU-C1 Finding There is no documented process for tracking cumulative impacts in PSA-08.

(This F&O originated from SR MU-C1)

The PRA model may not be representative of the as-built, as-operated plant.

Recommend adding guidance to PSA-08 for documenting the impact of individual changes and tracking the cumulative impact of multiple changes.

Procedure change with no fire PRA impact. The fire PRA is based on Internal Events model updated following last refueling outage.

Procedure has been updated to add cumulative impacts to the tracking list. This issue resolved.

2-01 IE-A4 Finding CN-RAM-14-030, section 7.1.1 documents the review of generic initiators included in NUREG/CR-3862. However, the CN-RAM-14-030, section 5.5 states: "Three pressurizer power-operated relief valves (PORVs) and three safety relief valves (SRVs) per plant are assumed to be a representative number for the industry." This is not a correct assumption as many plants only have two PORVs and not three. Additionally, application of the SRV failure rates to the PORVs may be an overly conservative approach.

(This F&O originated from SR IE-A4.)

The use of applicable generic initiator rates helps in the development of technically adequate plant initiator frequencies.

Consider creating a more accurate assumption that accounts for the number of units with 2 PORVs vs 3 PORVs. Also, reconsider the applicability and appropriateness of the failure mode data used; e.g., does the data represent truly random failure of an SRV/PORV or was it the result of a small leak or a pressure challenge where the valve failed to re-seat?

SR Met at CC-I/II.

No impact on Fire PRA since Fire PRA does not use internal events PRA Initiating Event frequencies.

The data use is not expected to have a significant impact on the IE frequency. As stated in the F&O it is conservative application. The internal events PRA CDF and LERF are slightly conservative based on this F&O.

2-02 IE-A5 Finding 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.)

Model completeness issue.

Per SR IE-A5, recommend performing 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. Include initiating events resulting from multiple failures if the equipment failures result from a common cause, and from routine system alignments as required by SR IE-A6.

The special 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 systems as documented in calculations DC00300-013 Rev. 1, IPE Initiating Event Notebook and DC00300-032, Rev. 1, IPE Special Initiator Event Tree Notebook. Calculation DC00300-013 was Referenced in the updated initiating event calculation but was not identified for the systematic review for special initiating events.

Calculations DC00300-013 and DC00300-032 were not provided to the Peer Review team and were not listed on the Supporting Requirement road map resulting in the F&O. The special initiating event list stood unchallenged from 1993 to 2016. This is a documentation and Supporting Requirement road map issue that does not impact Fire PRA.

2-03 IE-D1 Finding Within the SR mapping tables in the various analyses, the N/A entries should point to documentation that discusses why the SR is N/A.

(This F&O originated from SR IE-D1)

Provides scrutability.

Within the SR mapping tables, document why an SR is N/A or point to such documentation.

The SR mapping tables did not point to appropriate calculations for IE analysis. The update initiating event calculation did not point to existing documents for Supporting Requirements. This is a documentation and Supporting Requirement road map issue this does not impact Fire PRA.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 4 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 2-05 IE-A8 Finding 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)

Model completeness issue.

Interview plant personnel (e.g., operations, maintenance, engineering, safety analysis) to determine if potential initiating events have been overlooked.

The F&O provided no specific concern that initiating events were missed. This is documentation of interviews. The PRA model development including initiating events was led by a licensed operator. Closure of this F&O is not expected to add any new initiating events.

The Fire PRA conducted and documented interviews. The Fire PRA also does not use the initiating events from the internal events model.

There is no impact on the Fire PRA.

2-06 IE-D3 Finding Not all of the assumptions are documented in the Assumptions section (e.g., pressurizer relief valve sticks open if SI is not secured following a secondary side line break, many of the ISLOCA assumptions).

(This F&O originated from SR IE-D3)

Facilitates verification of the analysis and model uncertainty analysis.

Recommend compiling all assumptions in the Assumptions section (e.g., pressurizer relief valve sticks open if SI is not secured following a secondary side line break, many of the ISLOCA assumptions).

The SR was met at CC-III.

This is a documentation issue with no impact on the Fire PRA.

Several items in the text of the calculation were identified as assumptions, but not specifically listed in the Assumption section of the calculation. No impact on Internal Events PRA quantification.

2-08 IE-C1 Finding 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.

Also, the approach documented in CN-RAM-14-030, section 5 not to update the LOCA initiators because that would have a significant impact on CDF is not technically valid.

(This F&O originated from SR IE-C1)

Generic data used in the development of initiator frequencies needs to be up-to-date and applicable.

NUREG/CR-6928 data provides industry initiator failure rates through 2013. Recommend using this document as a more up-to-date data source. Revise the approach taken to quantify LOCA initiators.

The Fire PRA has its own initiating events frequencies and is not impacted by the Internal Events model in this regard.

A comparison of VCSNS LOCA IE frequency with the 2015 Parameter Estimation Update is as follows; Model 2015 LLO 1.66E-6 5.91E-6 MLO 4.00E-5 1.50E-4 SLO 5.00E-4 4.01E-4 The Medium LOCA (MLO) is higher. The value comes from expert elicitation in NUREG-1829.

The stated concern leading to the increase is Primary Water Stress Corrosion Cracking (PWSCC) which has been significantly reduced at VCSNS with the new reactor vessel head and weld overlays for the pressurizer nozzles. Also, Medium LOCA represents approximately 1.3% of current internal events CDF while Small LOCA (SLO) is about 20.5%. Reduction in SLO offsets the increase in MLO.

Closure of this F&O may result in minor increase in Internal Events CDF and LERF due primarily to the increase in Medium LOCA frequency.

2-09 IE-C2 Finding 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)

The use of more recent plant-specific data is required to help assure that calculated initiator frequencies are reflective of current plant experience.

Perform a review of more recent operating experience using guidance in PSA-08 step 6.2.3.A to help assure that the list of initiating events is complete. Retain information on all plant experience reviewed, not just the list of LERs for events that were kept for quantification input. This resolution can be used to meet SRs IE-A7 and IE-C2.

The Fire PRA has its own initiating events frequencies and is not impacted by the Internal Events model in this regard.

From 2008 through 2015 there were 29 VCSNS LERs. Over this 8 year period, 4 are at power PRA related initiating events. Closure of this F&O is expected to have a negligible impact on CDF and LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 5 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 2-10 IE-C5 Finding1 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.)

To develop plant-specific initiator estimates, an up-to-date, plant-specific availability factor should be developed and applied.

Recommend developing and using a plant-specific availability factor based on recent data.

The industry availability factor was applied to industry data to adjust for criticality years. This value was then Bayesian updated with plant specific data. A plant specific availability factor was used for the plant specific applications. This issue is resolved. The Fire PRA is using the appropriate data for plant specific availability factor.

2-11 IE-C6 Finding Loss of an AC bus, failure of the chilled water system, and consequential LOOP (post trip, after another initiating event occurs) are missing from the documentation or were screened using an inadequate basis. For example, in section 7.2.4.4, for loss of an AC bus the criteria listed does not discuss the impact of having to administratively shut the plant down within a short time frame due to LCO requirements with an entire electrical bus out of service.

(This F&O originated from SR IE-C6.)

Model completeness issue.

Recommend more clearly documenting each of the missing potential initiating events and add or screen them as appropriate. Ensure that screening meets the requirements of IE-C6 (if used) and document appropriately.

The Fire PRA has its own initiating events frequencies and is not impacted by the Internal Events model in this regard.

The initiating events listed were or could have been screened based on plant specific considerations. These include:

VCSNS has three chilled water packages (one for each train and a swing chiller). Rapid shutdown due to loss of chilled water is unlikely. Additionally, Loss of HVAC calculations support equipment mission times with no chilled water available.

Summer has the ability to feed each ESF Essential bus from either of two ESF offsite power supplies. Loss of offsite power to a single bus would not result in rapid plant shutdown.

With current operating practices, a loss of VCSNS generation is very unlikely to result in a loss of the electrical grid.

Therefore, this F&O will have minimal impact of the internal events model.

2-12 IE-C9 Finding CN-RAM-14-030 does not document the development of the fault tree initiator models or the quantification of the SSIEs.

(This F&O originated from SR IE-C9.)

Absent the documentation of the SSIE, it is difficult to verify the completeness of the model or compare SSIE frequencies to those of similar plants.

Recommend revising CN-RAM-14-030 to document the development of the fault tree initiator models and the quantification of the SSIEs.

The Fire PRA has its own initiating events frequencies and is not impacted by the Internal Events model in this regard.

The IE calculation CN-RAM-14-030 References the system notebooks for the fault tree SSIE development.

The Peer Review team was not supplied with the appropriate calculations which have the fault tree development for SSIE. The updated system notebooks Reference to an early revision for the actual fault tree development.

No significant impact to the internal events model is expected from closure of this F&O.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 6 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 2-13 IE-C14 Finding ISLOCA analysis performed in CN-RAM-14-030 documented the applicable ISLOCA pathways based on the plant features and procedures in accordance with WCAP-17154-P; see Appendix B of CN-RAM-14-030. However, the treatment of ISLOCA pathway 410-S appears to be overly conservative.

(This F&O originated from SR IE-C14.)

Over-conservatism in the ISLOCA analysis can lead to over-estimation of LERF.

Recommend reviewing the analysis of ISLOCA pathway 410-S to assure that it is a valid pathway and is not being quantified in an overly conservative fashion.

The SR is met at CC-I/II.

The ISLOCA pathway 410-S is not modeled in Fire PRA since an MSO cannot cause this ISLOCA scenario.

The 410-S ISLOCA is not a significant cutset for the Internal Events model with initiating event frequency of 2.13E-10/yr and percent of CDF of 0% (<0.1%). Impact on Internal Events CDF and LERF is minimal.

2-14 IE-C15 Finding No frequencies are quantified for SSIEs, which are embedded in the system fault trees. For almost all initiators, EFs are assumed rather than being based on the data.

(This F&O originated from SR IE-C15.)

Impacts the uncertainty bands associated with CDF and LERF.

Recommend deriving error factors based on the characteristics of the initiator distribution rather than using an assumed error factor.

Error Factors for the internal events model do not impact the results of the Fire PRA.

Error Factors are used for statistical analysis based on assumed distribution characteristics.

The Error Factors do not impact the internal events model CDF or LERF values.

2-15 HR-A1 Finding CN-RAM-14-033, Table 7.1-1 contains a list of procedures that were reviewed, organized by system. However, the list does not cover all modeled systems. For example, no procedures for Electric Power, RBCU, HVAC or the Pressurizer Pressure Relief System are included. There should be some discussion since the system notebooks point back to the HRA to justify that there are no pre-initiators for those systems.

(This F&O originated from SR HR-A1.)

This is a potential model completeness issue.

Recommend reviewing and ensuring that all PRA-modeled system procedures are evaluated for the potential to result in pre-initiator equipment unavailability due to maintenance or system alignment. Adhere more rigorously to industry screening criteria guidance

[NUREG-CR-4772 (ASEP); NUREG-1792, section 4.2.3.1].

This SR is met at CC-I-III The example systems would not affect Type A human error modeling or are reviewed with other system maintenance.

Electrical power systems out of service would be obvious because of equipment not starting during post-maintenance testing or train swaps and indicating lights not being lit in the control room.

The possibility for reactor building cooling unit (RBCU) unavailability is covered in the review of service water procedures.

HVAC is not modeled in the VCSNS PRA.

This issue is related to improving documentation to provide a clear path to SR documentation.

2-17 HR-B1 Finding Procedures appear to have been screened from further review at too high a level based on known faulty screening criteria as discussed in the Peer Review lessons learned document, using bases such as: Assume post maintenance testing would reveal any errors, Includes verified restoration line-up, or Unlikely to render the system inoperable. NUREG-1792, section 4.2.3.1, which was used as a basis for screening, requires that there be a compelling signal "(e.g., annunciator or indication) of improper equipment status or inoperability in the control room, it is checked at least once per shift or once per day, and realignment can be easily accomplished." Also, in some cases the screening process appears not to have been consistently applied. Actions FBVCC------HE, OACSAVXVG8153FC and OACSAVXVG8154FC were not screened even though they were related to SOP procedure while all other SOP procedures in the table were screened.

(This F&O originated from SR HR-B1.)

Inadequate screening of pre-initiator human interactions can result in an underestimation of risk due to equipment realignment.

Recommend adhering more rigorously to industry screening criteria guidance [NUREG-CR-4772 (ASEP);

NUREG-1792, section 4.2.3.1]. Ensure pre-initiators are screened consistently.

The comments in Table 7.1-1 are not fully explanatory for example as to why a particular IC procedure is unlikely to render the system inoperable. However, the conclusions are expected to remain valid. This is a documentation issue that will not result to significant changes in the Type A human reliability list when resolved.

Resolution of this F&O is not expected to significantly affect the internal events PRA or the Fire PRA.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 7 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 2-18 HR-F2 Finding Examples were found where the Time window for success is not based on AS/SC based timings, (e.g. OA_AAC_SBO, an assumption was used in lieu of an actual AS timing value which should be available. Additionally, for other HFE the limiting timing for which it is applied was not used, e.g. BCPM--

XPP39CHE is used for both LOOP and LCCW scenarios but the limiting timing was not selected).

(This F&O originated from SR HR-F2.)

The bases for the HRA timing information need to be technically sound.

Recommend reviewing and adjusting timings for the HFEs and use available AS/SC timings. For example, OA_AAC_ABO should be based on the limiting RCP seal leakage value and the BCPM--XPP39CHE should use the more limiting LOOP values as the EDG will over speed trip shortly after starting (if loaded or slightly longer if unloaded) or split the actions.

The SR was met at CC-III.

For the Internal Events PRA model, the basis for the timing of a few HRA was questioned. There may be small changes to the HFE values for these items for the Internal Events model but no impact on Fire PRA.

For the specific cases mentioned in the F&O:

OA_AAC_SBO - the alternate AC power supply was designed to be available in 60 minutes, so this was used as the total time required to align. The RCP seal LOCA is also 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, so there is no impact to the HRA.

For alignment of Service Water BCPM--

XPP39CHE, the total time available is consistent for either LOOP or LCCW cases.

The EDG overspeed postulated in the proposed F&O resolution is not consistent with VCSNS EDG experience. CDF or LERF will not increase for internal events or Fire PRA once this F&O is resolved.

2-19 HR-G4 Finding Certain instances were identified where the timing was not always used or bounded appropriately for the action. The point in time at which the cue is discussed and can be reasonably understood as to the basis but should be enhanced. Further instances were found where the T-0 was set to the time of the cue (Tsw was adjusted accordingly) and it should remain at the time of the event (trip). This adjustment can impact the dependency negatively.

(This F&O originated from SR HR-G4. Also, see F&O 2-18.)

The bases for the HRA timing information need to be technically sound.

Recommend basing the time available for completing actions on realistic T/H analysis.

The Fire PRA adjusted the delay times for these items, so there is no impact for Fire Specific HEPs.

An evaluation was completed using the EPRI HRA calculator for the dependency analysis to conservatively estimate the impact on the internal events PRA. The HRA calculator was used with existing timing with no credit for factors such as intervening success. The CDF increased by approximately The CDF increased by approximately 1.9E-6 per year and LERF by approximately 3.8E-8 per year. These increases are within margin for RG 1.174.

2-20 HR-G5 Finding For the majority of the HFEs reviewed the execution time stated that it was estimated. Execution time should be based on a combination of operator interviews and a standard guidance for operator execution times per step in the EOPs/AOPs/etc.

Industry guidance from most similar PWRs is on the order of 30 seconds to 1 minute per step based on plant-specific operator knowledge documents.

(This F&O originated from SR HR-G5.)

The bases for the HRA timing information need to be technically sound.

Recommend developing or verifying that operator guidance exists for expected timing on EOP/AOP/etc.

steps or at a minimum it should be acceptable to use such guidance from a similar plant adapted to plant-specific training features and simulator observations.

Operator interviews were conducted to support the time steps. This documentation was from an early HRA analysis that was not clearly Referenced in the update HRA analysis. The interview documentation was not provided to the Peer Review team. The traceability of the documentation needs to be updated, but there is no impact on the HRA values. There is no impact on the Fire PRA.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 8 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 2-21 HR-G6 Finding Based on a review of the HFEs, it is suggested to perform a sensitivity analysis on HFEs that have short windows for success (~15 min or less) using the HCR\\ORE method.

(This F&O originated from SR HR-G6.)

As the difference between time available and time required decreases, the failure probability increases and can surpass that derived using the Caused-Based approach.

For HFEs that have short windows for success (~15 min or less), recommend performing a sensitivity on the HEP quantification method used (HCR/ORE vs. Cause-Based) and document the insights.

SR is met at CC-III This is a documentation issue with no impact on the Fire PRA.

Early analysis was completed using HCR/ORE.

Original F&O HR-08 from the 2002 peer review made a similar comment. This issue was addressed in 2003. CBDT/THERP was used and resulted in higher numbers so CBDT/THERP method is being conservatively used (instead of HCR-ORE). Fully addressing the current F&O is unlikely to show that either CDF or LERF is non-conservative.

2-22 HR-H1 Finding Some risk-significant operator recovery actions are not currently modeled in the internal events model. Such events should be included as they could significantly impact the results.

Examples include an operator action to align recirculation for the RHR system and close the PORV block valve after the PORV has been challenged, or terminate SI after the SRV has been challenged, and to manually align equipment following a failure of ESFAS. Action OAESF1/OAESF2/OAESF3 were provided by VCS but did not appear in the modeled logic and should be included as it allows operators to start equipment after a failure as directed by procedures.

(This F&O originated from SR HR-H1.)

Omission of expected operator recovery actions can result in overly conservative risk estimates.

Recommend incorporating documented recovery actions into the model.

The Fire PRA has done extensive reviews to address conservatisms to reduce CDF and LERF.

There is no impact on the Fire PRA.

For the Internal Events PRA model, the updates will reduce CDF and LERF by a small amount.

The Internal Events PRA model is conservative with regard to this F&O.

4-21 LE-C11 Finding It was assumed that early containment failures result in core damage. Therefore, credit for equipment survivability after containment failure has not been taken.

(This F&O originated from SR LE-C11)

By not taking credit of containment spray/recirculation spray or any operator actions after containment failure, utility only meets CAT I of this SR, per its stated intent.

Review significant accident progression sequences resulting large early release to determine if credit can be taken for equipment operation or operator actions after containment failure that could reduce LERF. Take credit for containment spray/recirculation spray or any operator actions after containment failure and justify it.

Both fire and internal events LERF models are dominated by containment bypass events rather than containment failure.

For the Internal Events PRA and the Fire PRA model, credit for containment spray or operator action may reduce LERF by a small amount.

Closure of this F&O is not expected to significantly impact LERF.

3-02 DA-A2 Finding Many undefined SSC boundaries in Table 6-2 of CN-RAM 034.

(This F&O originated from SR DA-A2)

It is assumed that the "N/A"s just mean it's not defined in 6928 and should just include the component itself, this should be explicitly stated.

Recommend ensuring that all SSC boundaries for all components are defined.

The Fire PRA is not impacted by the internal events model documentation.

For the internal events model, the Peer Review team was not provided the appropriate documentation to confirm the SR is met.

Component and system boundaries for the VCSNS PRA are defined in the IPE Support State Model Notebook and the Dependency Notebook.

3-03 DA-A3 Finding Table 7.2.1-1 of CN-RAM-14-005 contains type codes with lognormal distributions and error factors of zero. This is nonsensical.

(This F&O originated from SR DA-A3)

May be using incorrect or inappropriate data for various type codes.

Recommend ensuring and confirming the correctness, validity, and appropriateness of failure distributions including uncertainty parameters.

The uncertainty work for internal events does not impact the Fire PRA.

Error Factors are used for statistical analysis based on assumed distribution characteristics.

The Error Factors do not impact the internal events model CDF or LERF values.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 9 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 3-06 DA-A2 Finding The EF system notes on page 39 that standby data was considered applicable for all three EF pumps. However, the model only appears to include two failure modes for the TDPs:

(a) Standby TDP fails to start and (b) Standby TDP fails to run.

This seems inconsistent with the NUREG/CR-6928 data set for standby TDPs which have three FMs:(a) Standby TDP fails to start, (b) Standby TDP fails to run less than 1 hr and (c) Standby TDP fails to run after 1 hr.

A cursory review if the Data Notebook indicates that this also applies to other standby components such as EDGs, MDPs, Fans and Compressors.

(This F&O originated from SR DA-C1)

Incomplete failure data being used.

Include both FTR>1 hr and FTR<=1 hr in failure data for FTR when split up in 6928 but not split up in plant-specific type codes.

The component failures are modeled as fail to start (FTS) and fail to run (FTR). The Peer Reviewer opinion/position was that FTR had to be broken into two basic events for FTR < 1 hr and FTR > 1 hr. This is not an ASME Standard requirement. Plant specific data has not been collected to support a transition to FTR < 1 hr and FTR > 1 hr. The model is set up consistent with the plant specific data source.

Some failure rate data for fail to run was based on older source data (EPRI TR-016780-V3R8, ALWR Utility Requirements Document) leading to the Peer Review comments. Incorporating more recent data from NUREG-6928 and 2015 Parameter Estimation Updates may have a small impact on CDF and LERF.

3-07 DA-B2 Finding 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)

May be inappropriately grouping components.

Recommend discussion that addresses accounting for outliers. Consider distinctions such as frequency of manipulation.

PRA failure rates are per hour or per demand. Per demand failure rates already take into account frequency of manipulation. Identifying seldom operated valves as "outliers" will not change reliability values.

3-10 DA-D8 Finding The use of old data is not limited. Table 7.1.3-1 of CN-RAM 005 shows that recent plant-specific data is combined with old plant-specific data in order to do Bayesian update. DC00300-152 shows that recent Test and maintenance data is combined with old plant-specific data.

(This F&O originated from SR DA-D8)

Old data may not be applicable.

Use only recent plant-specific data or confirm that the old data is still applicable.

This is a documentation issue for the internal events model and does not impact the Fire PRA.

The data update calculation did not specifically justify the use of all plant specific (and particularly some of the older) data. Recent data indicates general improved equipment reliability and reduced unavailability. Using only recent data to close this F&O is expected to improve both CDF and LERF.

3-11 DA-D6 Finding No evidence that plant-specific failures were reviewed for common cause failures.

(This F&O originated from SR DA-D6)

Generic common cause failure probabilities may not apply to plant.

Review plant experience for common cause failures.

This is a documentation issue for the internal events model and does not impact the Fire PRA.

There have been no common cause failures documented at the plant. This was not clear in the documentation. No impact is expected for the internal events model.

3-12 DA-E1 Finding Many items in the ASME Standard Roadmap for CN-RAM 034 are labeled N/A and appear to be undocumented.

(This F&O originated from SR DA-E1)

Cannot verify the validity of analysis.

Add data and justifications from previous model to Data Update document.

This is a documentation issue for the internal events model and does not impact the Fire PRA.

Calculation CN-RAM-14-034 is the Data Notebook.

The road map covers the DA supporting requirements. The vendor calculation (CN-RAM-14-034) only Referenced documents generated by the vendor in the road map. A number of the SRs are supported by VCSNS plant calculations.

These calculations were identified and provided to the Peer Review team to complete the SR assessment. The road map documentation needs to include both the vendor References and the site documentation. Closure of this F&O is not expected to change the CDF or LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 10 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 3-15 DA-C5 Finding 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)

May be inaccurately counting failures.

Update data guideline (PSA-05) to address occurrences of multiple failures. This issue is an item in LTR-RAM 35.

This is a documentation issue for the internal events model and does not impact the Fire PRA.

The one repeat failure (EDG) was documented, but it was not clearly documented that there were no other repeat failures. The guideline was also not specific. No impact is expected for the internal events model.

The PRA guideline (VCS-PRAG-0004) has been updated to include discussion of repeated failures in short time intervals.

3-16 DA-C6 Finding There is no evidence that additional demands from post-maintenance testing were considered.

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

May be inaccurately counting demands.

The documentation could be enhanced by including an explanation that restoration demands were not included in the count. This issue is an item in LTR-RAM-16-35.

This is a documentation issue for the internal events model and does not impact the Fire PRA.

This was a documentation issue only, as these demands were considered and accounted for in the PRA models. The SR road map did not point to these calculations.

The PRA guideline (VCS-PRAG-0004) has been updated to include documenting the consideration of additional demands from post-maintenance.

3-22 DA-D4 Finding Only prior and posterior distribution mean values compared. No other distribution characteristics compared to ensure reasonableness.

(This F&O originated from SR DA-D4)

Prior data may not be applicable to plant.

Compare prior and posterior distribution characteristics.

This was a documentation issue only, as these comparisons were made and considered for reasonableness in the PRA models This finding does not impact the internal events or Fire PRA.

The PRA guideline (VCS-PRAG-0004) has been updated to include checking the results to ensure reasonableness.

4-01 SC-A5 Finding The current model utilizes mission times less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by the statement #1.

(This F&O originated from SR SC-A5)

The SR indicates that a minimum mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is required. This can be divided into phases, but the total must equal 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The success criterion notebook (CN-RAM-14-032) allows for success if the plant can be classified as being in hot standby (mode 3) where no further operator actions or system activations are required to mitigate the effects of the initiating event and the operators can proceed to either plant shutdown (mode 4 or 5) or return to power generation (mode 1). This could occur in a shorter period than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. There are also examples of mission times that are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> specified in the system notebooks.

Recommend revising all accident sequences and response systems to be consistent with a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time.

Mission times in the model are based on 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or a documented mission time. An example of a documented mission time is EFW pump operation for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during a small LOCA. EFW provides core cooling. Within this 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> the operators depressurize the RCS and/or aligned for long term cooling. The long term cooling then has a mission time of 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />. The Peer Review team did not fully agree with this approach and recommended changing all mission time to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If adjustments to mission times are found necessary, the closure of this F&O is not expected to have a significant impact on CDF or LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 11 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 4-04 SC-B3 Finding Thermal hydraulic analyses have been used to support the success criteria. However, many issues were identified with success criteria. For example, In an SGTR event with failure of SG isolation (SGI) and RWST Refill (RF), core damage is avoided if (long-term heat removal) LTHR is successful. Its not clear how success is achieved if all this function includes is steam relief. Or is it assumed that this function is similar to Cooldown function? Is there a MAAP analysis to support this success criteria? Another example, bleed and feed success criteria MAAP case assumes the action is initiated at the time of the cue (12% wide range level) which allows no time for the operator action. With the extra time used before the action is actually initiated, it is not clear whether the case would still be successful with 1 Charging pump and 1 PORV. Yet another example, Assumption 3 in Section 5.0 states that 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> was used as a mission time for emergency feedwater in an SLOCA event since it is used for RCS cooldown and depressurization only if HPI fails. Looking at the SLOCA event tree and fault tree sequence SLO-7, there is no mitigation if HPI fails (core damage occurs).

(This F&O originated from SR SC-B3)

Thermal hydraulic analyses have been used to support the success criteria. However, many issues were identified with success criteria. For example, In an SGTR event with failure of SG isolation (SGI) and RWST Refill (RF), core damage is avoided if (long-term heat removal)

LTHR is successful. Its not clear how success is achieved if all this function includes is steam relief, or if it is it assumed that this function is similar to a cooldown function. There is no clearly identified MAAP run to support this success criteria. In one another case, a feed and bleed success criteria MAAP run assumes the action is initiated at the time of the cue (12% wide range level) which allows no time for the operator action. With the extra time used before the action is actually initiated, it is not clear whether the case would still be successful with 1 Charging pump and 1 PORV.

In yet another example, Assumption 3 in Section 5.0 states that 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> was used as a mission time for emergency feedwater in a SLOCA event since it is used for RCS cooldown and depressurization only if HPI fails. Looking at the SLOCA event tree and fault tree sequence SLO-7, there is no mitigation if HPI fails (core damage occurs).

Recommend performing of a systematic review of success criteria to better understand the PRA model, and success criteria used in the model, and correction of any errors or better justification of applicability. Also recommend that the acceptability of generic analysis be justified based on consistency with the model and as-built as-operated plant, as well as any relevant plant procedures.

The Success Criteria have been updated over time in several Reference documents. Based on the F&Os, these documents need to be more clearly organized and the criteria directly Referenced to the analysis. The specific items identified in the peer review are addressed as follows; The SGTR success criteria came directly from WCAP-15955. The Long Term Heat Removal (LTHR) includes an aggressive cooldown to 200 F.

The Feed and Bleed for SGTR is covered in CN-RAM-14-041, Section 5.2.3. The operator action is bounded by small LOCA MAAP run, except for longer time for SG dryout. The operator action (OAB1) is modeled as 15 minutes per CN-RAM-14-033.

The Small Break LOCA discussion in the Success Criteria Notebook was not correct.

There is no mitigation modeled if HPI fails.

The basis for this is documented in a MAAP run which shows RCS depressurization and low pressure injection/recirculation without HPI is time limited and not completed prior to core temperatures exceeding 1200F. The event tree is correct. There is no mitigation if HPI fails.

The specific concerns listed do not impact CDF or LERF. Closure of this F&O is not expected to have significant impact CDF or LERF.

4-05 SC-B4 Finding Utility used mixture of MAAP5 runs and control room simulation runs to support their success criteria, HRA timing and any assumptions they used in PRA. While MAAP5 has been widely used in PRA and is completely appropriate, the use of the operator simulations runs are questionable, especially with no documentation available to review the inputs and outputs, and no maintenance program for maintenance in accordance with the PRA Standard (for example, has it been benchmarked against other software and has it been used within its known limits of applicability).

(This F&O originated from SR SC-B4)

A mixture of MAAP5 runs and control room simulator runs to support the success criteria, HRA time windows and any assumptions they used in PRA. While MAAP5 has been widely used in PRA and is completely appropriate, the use of the operator simulator runs is questionable, especially without supporting documentation available to review the inputs and outputs, and no identified maintenance program in accordance with the PRA Standard (for example, has it been benchmarked against other software and has it been used within its known limits of applicability).

Recommend using MAAP5 to support all success criteria and HRA time windows, and document all cases in PRA documentation with a clear cross-Reference between the thermal-hydraulic cases and the success criteria/HRA time windows they support.

The Peer Reviewer opinion was that simulator timing could not be used to support HRA timing.

Simulator timing has been accepted in the industry to support HRA. Closure of this F&O is not expected to significantly impact CDF or LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 12 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 4-06 SC-B5 Finding The only comparison performed was for a SGTR event, with no results shown. This is not enough to demonstrate reasonableness and acceptability of the results.

(This F&O originated from SR SC-B5)

The only comparison performed was for a SGTR event, with no results shown. This is not enough to demonstrate reasonableness and acceptability of the results.

Recommend performing additional thermal-hydraulic runs for comparison purposes to demonstrate reasonableness and acceptability of the results.

The Success Criteria notebook contained the SGTR event comparison and the SR road map pointed to this comparison. A separate document provided for SR QU-D4 provides a PRA model comparison study including Success Criteria for all initiating events. QU-D4 was met at CC-II/III. The following statement was made by the peer review team for QU-D4 LTR-RAM-16-12-2 documents a comparison of VCS initiator frequencies, success criteria, and CDF and LERF results to those from Shearon Harris Nuclear Power Plant and Joseph M. Farley Nuclear Plant Units 1 and 2. Except for the ISLOCA frequency difference, the causes for significant differences are investigated. The available documentation was not properly mapped to SR SC-B5. Closure of this F&O is not expected to impact CDF or LERF.

4-07 SC-C1 Finding The documentation success criteria does not facilitate the peer review since navigation in the notebooks is a real challenge due to lack of cross referencing.

(This F&O originated from SR SC-C1)

The documentation of the success criteria analysis does not facilitate applications, upgrades and peer reviews since navigation within the documentation is challenging due to that lack of adequate cross-referencing.

Recommended providing clear cross-References between the thermal-hydraulic runs and related success criteria/HRA time windows.

This is a document issue with the internal events model cross-referencing Success Criteria with supporting analysis. The Success Criteria Notebook is largely in text format with supporting documents Referenced throughout. The Peer Reviewer position was that the specific thermal-hydraulic analysis in the Referenced document should be listed in the Success Criteria Notebook.

This change would facilitate the peer review at a higher level. Closure of this F&O is not expected to impact internal events CDF or LERF.

4-08 SC-C2 Finding Most of the items in this SR are not met. For example, identification of calculation and what they support, a description of limitations, basis for establishing time window for operator actions, etc.

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

Most of the items in this SR are not met. For example, identification of calculation and what they support, a description of limitations, basis for establishing time windows for operator actions, etc.

Recommend explicit documentation of all items identified in this SR. Also recommend clear cross-References between the thermal-hydraulic analysis runs and related success criteria/HRA time windows.

This is a document issue with the internal events model cross-referencing Success Criteria with supporting analysis. The Success Criteria Notebook is largely in text format with supporting documents Referenced throughout. The Peer Reviewer position was that the specific thermal-hydraulic analysis in the Referenced document should be listed in the Success Criteria Notebook.

This change would facilitate the peer review at a higher level. Closure of this F&O is not expected to impact internal events CDF or LERF.

4-11 LE-A4 Finding The LE-A1 and LE-A2 characteristics are accounted for using different methods including treatment in Level 2 CETs, bridge trees, PDS. One finding was identified.

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

Accident sequences with RCP seal LOCA of 480 gpm have been assumed to depressurize enough to be considered unlikely to cause ISGTR (see notebook CN-RAM-14-035, Section 5, Assumption 19). This was done based on WCAP-16341 guidance. No severe accident/MAAP analysis exists to demonstrate this depressurization. It should be noted though that RCP seal LOCA results in loop seal clearing and thus significantly increases ISGTR probability. Therefore, the assumption that catastrophic seal LOCA occurs very early in the accident versus at the time of core damage is non-conservative from LERF standpoint.

Consider modeling accident sequences with RCP seal LOCA contributing to LERF, or justify with severe accident calculations that enough RCS depressurization will take place to avert TI-SGTR.

The SR is met at CC-III.

The F&O on Internal Events LERF is limited to the one scenarios for RCP seal LOCA. The analysis followed the application in WCAP-16341 which is an accepted method. The Peer Reviewer had separate experience regarding the modeling of the event. Final resolution of the F&O is not expected to result in significant impact of LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 13 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 4-13 LE-B1 Finding All LERF contributors from the table have been considered in LERF analysis. However, in-vessel injection was not credited.

(This F&O originated from SR LE-B1)

No credit was taken for in-vessel recovery to stop core-melt process and prevent LER.

Recommend modeling in-vessel recovery to prevent LERF, or justify not modeling by demonstrating its low likelihood of success or insignificant impact on LERF.

The SR is met at CC-II.

No credit it taken for in-vessel recovery after core damage. The offsite power model is not adequately detailed to model recovery between core damage and vessel breach. The internal events is slightly conservative is this matter. For the Fire PRA this option is not available, so there is no impact.

4-14 LE-C1 Finding No severe accident analyses exist to support determination on whether or not an accident sequence should be binned to LERF.

(This F&O originated from SR LE-C1)

Plant-specific severe accident analyses (for example, with MAAP) are needed to determine whether an accident progression sequence results in LER or not.

Recommend performing severe accident analyses to support accident sequence binning.

The simplified Level 2 Analysis is based on WCAP-16341-P. WCAP-16341-P is an extension of NUREG/CR-6595 and includes realistic quantification of containment threats resulting from high pressure failure of the reactor vessel and additional detail on the treatment of Interfacing System LOCA (ISLOCA) and Induced Steam Generator Tube Rupture (I-SGTR).

SR LE-C1 Capability Category II states JUSTIFY any generic or plant-specific calculations or References used to categorize releases as non-LERF contributors based on release magnitude or timing. NUREG/CR-6595, App. A provides a discussion and examples of LERF source.

WCAP-16341 supports binning to LERF on a generic basis.

See F&O 4-23 for discussion on application of generic basis. Closure of this FO is not expected to significantly impact LERF.

4-15 LE-C2 Finding SAMG operator action to depressurize RCS has been credited to prevent LER. However, there is a finding identified for this SR.

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

This SAMG action has been credited for both SBO and non-SBO sequences. It's not clear how the PORVs would be operated in an SBO event after DC battery depletion.

Recommend providing additional justification for PORV operability for SBO sequences with DC power depletion, or do not take credit for this action for SBO events.

The SR is met at CC-III One item was noted on RCS PORV use after battery depletion in SBO condition. The LERF model follows WCAP-16341 and models early RCS depressurization with the RCS PORVs.

Timing of the sequence is not specified. Closure of this F&O is expected to have little impact on LERF.

4-16 LE-C4 Finding Model logic is included for realistic estimation of accident progression sequences resulting in a LER. This is documented in notebook CN-RAM-14-035, Section 6.0 and Appendix A. A finding is identified.

(This F&O originated from SR LE-C4)

Plant-specific severe accident analyses (for example, with MAAP) are needed to demonstrate effectiveness of any operator actions and fission product scrubbing.

Recommend performing severe accident analyses to demonstrate that any operator actions and fission product scrubbing credited to reduce LERF are effective.

This SR is met at CC-II.

Generic calculations were used to demonstrate the effectiveness of the modeled operator action and effectiveness of fission product scrubbing.

Additional plant specific calculations may slightly impact LERF results, but were not deemed necessary to meet CC-II.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 14 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 4-17 LE-C5 Finding Realistic or plant specific analyses have not been performed for system success criteria for significant accident progression sequences that result in LER.

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

Realistic or plant specific analyses have not been performed for system success criteria for significant accident progression sequences that result in LER. For example, no analyses exists to determine success criteria of SAMG action to depressurize RCS (whether one or two PORVs are necessary).

Recommend performing plant specific analyses (for example, using MAAP) to support system and operator action success criteria used in the LERF model.

The LERF effort was completed based on a realistic generic analysis for system success criteria. This is acceptable for Capability Category II which states USE appropriate realistic or plant-specific analysis for system success criteria for significant accident progression sequences. The basis for applicability of the generic analysis is being expanded in the Level II notebook.

Resolution of this F&O is not expected to significantly impact LERF.

Subsequent to the Peer Review a plant specific station blackout analysis was performed for SBO\\PI-SGTR. The MAAP run addressed treatment of the main steam safety valves. This addressed the difference between the Reference plant [NUREF-1570] and plant specific design.

4-18 LE-C3 Finding No review of significant accident progression sequences contributing to LERF could be found to determine if repair can be credited to reduce LERF.

(This F&O originated from SR LE-C3)

To meet the intent of CAT II of this SR, a review of significant accident progression sequences contributing to LERF is needed to determine if repair can be credited to reduce LERF.

Recommend review of significant accident progression sequences contributing to LERF to determine if repair can be credited to reduce LERF. If, after review, it is decided that no repair is to be credited, then document it.

The Fire PRA has completed reviews of top cut sets for LERF.

The internal events review of the top 100 LERF cutsets was performed. Repair actions were identified as a part of the review. These have the potential to lower LERF. The LERF model is conservative in this regard.

4-19 LE-C7 Finding One operator action has been credited, SAMG action to depressurize RCS. See F&O on LE-C2.

(This F&O originated from SR LE-C7)

This SAMG action has been credited for both SBO and non-SBO sequences. It's not clear how the PORVs would be operated in an SBO event after DC battery depletion.

Recommend providing justification for PORV operability for SBO sequences with DC power depletion, or do not take credit for this action for SBO events.

The SR is met at CC-III.

There is no impact on the Fire PRA since it does not use a recovery action to open RCS PORV after core damage.

For the internal events PRA, one item was note on RCS PORV use after battery depletion in SBO condition. The LERF model follows WCAP-16341 and models early RCS depressurization with the RCS PORVs. Timing of the sequence is not specified. Closure of this F&O is expected to have little impact on LERF.

4-20 LE-C9 Finding Pressurizer PORVs have been credited post-core damage for RCS depressurization. No justification is provided for PORV survivability in harsh environment inside containment after core damage.

(This F&O originated from SR LE-C9)

Pressurizer PORVs have been credited post-core damage for RCS depressurization. No justification is provided for PORV survivability in harsh environment inside containment after core damage.

Recommend providing justification for PORV survivability in harsh environment inside containment after core damage.

The LERF model was developed consistent with WCAP-16679 with regard to the PORV function.

The recommended modeling states (Section 2.13)

It is recommended that no failures of the pressurizer PORV due to thermal effects be modeled in the PRA for the base model (i.e.,

assume that the competing factors result in an even trade off).

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 15 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 4-23 LE-C13 LE-A5 Finding Containment bypass analysis is performed appropriately.

However, major issue identified with crediting scrubbing in SG. It is based on an assumption which may be non-realistic and non-conservative.

(This F&O originated from SR LE-C13)

SGTR sequences with successful EFW have been binned to small releases regardless whether the ruptured SG was isolated or not. Note that there is a good chance that isolation of the ruptured SG will cause SG dryout. No severe accident analysis exists to demonstrate that it will not. Also note that even if the isolation function #SGI is failed, it does not necessarily mean that EFW has not been isolated; it may just mean that isolation has failed from the main steam side of the ruptured SG, but has been isolated from the FW side.

Recommend performing severe accident calculations to demonstrate that these sequences do not result in SG dryout and large release, or bin these SGTR sequences to large releases.

This is only applicable to the Internal Events and does not impact Fire PRA.

The internal events model assumed that SGTR with emergency feedwater available would have sufficient scrubbing to preclude a LERF and be a Small early release. The Peer Reviewer challenged this assumption for SGTR conditions when the secondary side was not isolated (failed open PORV or safety). The concern was dry out of the S/G would result in no scrubbing and lead to LERF. An effort to address this F&O is in progress with DRAFT results available.

Review of the SGTR end states compared with the NRC State of the Art Reactor Consequence Study (SOARCA) was completed using a similar 3-loop PWR. Cases with high pressure injection pumps and emergency feedwater (and secondary side not isolated) are expected to screen to large late releases. The case without high pressure injection and emergency feedwater secured was modeled n MAAP 5.1. For this particular scenario S/G dryout was predicted and screened to LERF. The expected impact on LERF is approximately 1% of total LERF. Closeout of this F&O is not expected to result in a significant change to LERF.

4-24 LE-D2 Finding No evidence could be found on evaluation of the impact of containment seals, penetrations, hatches etc. on containment ultimate capacity.

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

To meet the intent of this SR, evaluation is required to justify no impact on containment performance/capacity. No evidence could be found on evaluation of the impact of containment seals, penetrations, hatches etc. on containment ultimate capacity.

Recommend providing evidence of evaluation of the impact of containment seals, penetrations, hatches etc.

on containment ultimate capacity. If no such evaluation exists, perform the evaluation and determine the impact on containment capacity or justify that it has no impact.

Design calculations for penetrations demonstrate they can withstand a pressure of at least 80.3 psig.

This is the same as the containment building. Test date for electrical penetration seals were completed at 92 psig for the limiting design.

Closure of this F&O is not expected to impact CDF or LERF.

4-25 LE-D3 Finding This assessment is based on VC Summer self-assessment.

Also, no analysis exists to determine if containment failure location has any impact on binning to LERF.

(This F&O originated from SR LE-D3)

No severe accident analyses exist for determination of impact of containment failure location on classification of any accident progression sequences as LER.

Recommend performing severe accident analyses to determine whether containment failure location has any impact on classification of any accident progression sequences as LER.

An effort to address this F&O is in progress with DRAFT results available.

The risk significant LERF sequences are primarily associated with bypass events and early core damage events without containment isolation.

These are applicable to above ground release points.

In the context of the VCS Level 2, the primary impact of this SR would be relevant for containment failure sequences where alternate release paths/locations with different consequences are possible. The early containment failure mode where this is theoretically possible is a consequence of an ex-vessel steam explosion. This would potentially fail the containment above or below ground. A review of steam explosion failure frequency suggests steam explosion failures are not risk significant and all ex-vessel steam explosions are treated as LERF. Changes in containment failure locations will not significantly impact LERF. Closure of this F&O will not significantly impact LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 16 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 4-26 LE-D5 Finding Secondary side isolation has not been modeled realistically for SGTR sequences since these were assumed not to result in LERF.

(This F&O originated from SR LE-D5)

This finding is related to finding on LE-A5.

Secondary side isolation needs to be considered realistically when determining whether an SGTR sequence results in LER.

Recommend performing realistic assessment of secondary side isolation in determination of classification of SGTR sequences as LERF contributors. This is technically already done in the Level 1 model. Need carry the Level 1 SG isolation logic to the Level 2 space for realistic assessment potentially supported by Level 2 severe accident calculations.

A detailed discussion on modeling of secondary side isolation and the impact of SGTR LERF is provided in response to F&O 4-23. Closeout of this F&O is not expected to result in a significant change to LERF.

4-27 LE-D6 Finding The analysis of thermally-induced SGTR was performed conservatively per WCAP-16341. The thermally-induced SGTR analysis was performed using event trees for SBO and non-SBO event trees. The analysis found to be too conservative to meet the intent of the CAT II requirement.

(This F&O originated from SR LE-D6)

The thermally-induced SGTR analysis was performed using event trees for SBO and non-SBO event trees. However, at the end, all end states leading to LERF were summed up to produce one conditional probability of LER for SBO, and one for non-SBO scenarios. Then, these values were applied to all respective accident sequences. For example, all high-dry transient sequences are assigned a LERF fraction of 0.335 which is too conservative. Factors such as RCP seal LOCA and SG depressurization can have significant impact on the conditional probability of LER.

Therefore, more detailed analysis is needed on accident sequence level to determine more realistic LERF contributions.

Recommend performing more detailed analysis on accident sequence level to determine more realistic LERF contributions. Either do not sum up the TISGTR event tree end state values, or apply the summed up value to more expanded Level 1 sequences to account for potential loop seal clearing and SG depressurization.

The LERF calculation has been updated to remove conservatism for a parallel PRA project (Seismic PRA model). This work resulted in a negligible impact on internal events LERF and net decrease in LERF for Seismic PRA.

These results are currently not incorporated in the Fire PRA model, but they are expected to have negligible or net decrease in LERF.

4-28 LE-E2 Finding Parameter estimates for characterization of accident progression phenomena were used in a conservative manner.

Examples are calculation of containment pressure due to hydrogen burn, containment failure due to steam explosion, containment bypass due to PISGTR and TISGTR, etc.

(This F&O originated from SR LE-E2)

To meet the intent of CAT II of this SR, more realistic parameter estimates should be used in characterization of accident sequence phenomena.

This is especially relevant for the top LERF cutsets that include steam explosion and PISGTR and TISGTR large early release contributors. In fact, steam explosion is not even included as a LERF contributor in Table 2-2.8-9 of ASME PRA Standard 2009 revision.

Recommend using more realistic parameter estimates for steam explosion and PISGTR and TISGTR values.

The LERF calculation has been updated to remove conservatism for a parallel PRA project (Seismic PRA model). The result is a negligible impact on internal events LERF and net decrease in LERF for Seismic PRA.

Incorporation of this work into the Fire PRA model, is expected to result in a negligible change or net decrease in LERF.

4-29 LE-F2 Finding LERF contributors were reviewed for reasonableness in self-assessment portion of notebook CN-RAM-14-035, Table D-2.

More detailed review is necessary to assure extra conservatism does not skew the results.

(This F&O originated from SR LE-F2)

The LERF contributors should be reviewed in more detail to make sure excessive conservatism has not skewed the results, and the LERF properly represents plant risk of large early release.

Recommend performing more detailed review of LERF contributors to make sure excessive conservatism has not skewed the results, and the LERF properly represents plant risk of large early release. Document the review within the body of the notebook, instead of the self-assessment.

The SR is met at CC-III.

The Peer review team indicated this F&O would not be issued, but it was published in the final report. The concern was that the LERF review was contained in the self-assessment of the model rather than the body of the calculation. There is minimal conservative impact on the internal events PRA model, and inclusion of this documentation in the calculation has no impact on the results of any of the VCSNS PRA models.

4-30 LE-F3 Finding Plant specific assumptions are identified and documented in Section 5 of notebook CN-RAM-14-035, and generic uncertainties are characterized in Section 7.4. In addition, certain sensitivity cases were performed and documented in Section 7.3. LERF aleatory uncertainty evaluation is performed with UNCERT in Section 7.2.2. One finding was identified on error factors.

(This F&O originated from SR LE-F3)

The two UNCERT evaluations were performed in section 7.2. to document the impact of the uncertainty of the Level 2 parameters. The first evaluation was performed assuming lognormal Error Factors (EFs) of 1.0 for all Level 2 BEs. The second evaluation was performed using best estimate EFs developed for this analysis. It is not clear whether there was any technical basis for the selection of the error factors in the UNCERT evaluations.

Recommend providing technical basis for selection of error factors in the UNCERT evaluations.

The SR was met at CC-III.

The basis for selection of the best estimate error factors is specifically covered in Section 7.2.2 of CN-RAM-14-035. The UNCERT evaluation was used to show that the aleatory uncertainty of the LERF Level 2 model is small. Resolution of this F&O is not expected to impact the internal events model LERF. There is no impact on the Fire PRA.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 17 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 4-31 LE-G2 Finding Documentation of LERF analysis was performed properly.

Detailed self-assessment is provided in Table D-2 of notebook CN-RAM-14-035.

(This F&O originated from SR LE-G2)

Some reviews of significant LER contributors were only documented in the self-assessment part of the notebook (for example, review for reasonableness of results). These reviews need to be documented in more detail within the main body of the notebook.

Recommend documentation of reviews of significant LERF contributors in the main body of the notebook.

The SR is met at CC-III Documentation of internal events LERF self-assessment has no impact on Fire PRA.

This is a documentation issue. Moving the LERF review to the main calculation will have no impact on LERF.

4-32 LE-G5 Finding Limits of applicability are documented in Section 8.3. However, this documentation appears to be minimum; more detailed documentation is needed to meet the intent of this SR.

(This F&O originated from SR LE-G5)

Limitations in VC Summer LERF analysis that would impact applications should be documented in detail.

Recommend documenting, in detail, the limitations in VC Summer LERF analysis that would impact applications.

For example, this is an at-power internal events model only, so therefore should not be used for any applications that require external events or low-power shutdown models.

The statement on limits of applicability of the internal events model has no impact on the Fire PRA.

This is a documentation only issue for internal events. Clarification of the document will have no impact on CDF or LERF.

4-33 LE-G6 Finding The quantitative definition of significant accident progression sequence is defined in notebook CN-RAM-14-035, Section 4.0.

However, it is not clear how it is used.

(This F&O originated from SR LE-G6)

Judging based on documentation of notebook CN-RAM-14-035, it is not clear how the definition of significant accident progression sequences was used. No significant accident progression sequences were identified.

For the purpose of reviewing significant accident progression sequences, recommend identifying them in the results per the definition used in the notebook Section 4.0.

This SR is met at CC-III.

This is a documentation concern with the internal events calculation statement that has no impact on the Fire PRA.

The Peer Review team wanted to see application of a defined term in the notebook. This is a documentation issue with no technical impact on the internal events PRA.

5-01 QU-B9 Finding Logic flags have not been set to TRUE or FALSE for all flags within the model. Additional cutsets have been generated in the final results that should not exist as they are non-minimal.

(This F&O originated from SR QU-B9)

Additional cutset are being generated in the results due to flag events remaining in the model that are not set to TRUE or FALSE, see cutsets 7 and 8 in the final CDF results where the only difference is a flag that is used for some sort of identification.

Recommend setting flags to either TRUE or FALSE, OR utilize a methodology whereby the quantifier (FTREX) can identify flag files. (Note: FTREX quantifier settings can be manipulated such that it can identify and ignore flag files during quantification and retain them in the results for the purposes of identification without impacting the results).

This item has been addressed in the subsequent internal events model revision. The issue is resolved with a net reduction in CDF.

The updated internal events model was used for the Fire PRA, so there is no impact on Fire PRA.

5-02 QU-A2 Finding The approach did identify a measurable change which was used to find an adjustment factor for SOKC. However, the analysis is believed to be incomplete.

(This F&O originated from SR QU-A3)

SOKC was not propagated throughout the common type codes. System level type codes exist such that the correlation does not propagate through correctly, e.g. MOV data for priors is the same for all systems but it is not being properly handled by UNCERT since it does not recognize the type codes are split.

Define common type codes for similar components and perform the UNCERT evaluation The SR is met at CC-III.

This item has been addressed in the subsequent internal events model revision. The issue is resolved.

The internal events model with the issue resolved was used to update the Fire PRA.

5-04 QU-B5 Finding Circular logic modeling was broken at the system notebook level and appears to be correct based on the existing system fault tree analysis. However, no guidance for performing circular logic breaking could be found in the quantification notebook, the systems notebooks, or the PSA-01 fault tree modeling guidance document.

(This F&O originated from SR QU-B5)

No guidance could be found for the breaking of circular logic in the quantification notebook, systems notebook, or the PSA-01 fault tree modeling guidance documents.

Recommend developing and documenting guidance for breaking circular logic.

The SR is met at CC-III.

The recommendation for developing a written process for breaking circular logic in the internal events model has no impact on the Fire PRA.

There will be no impact on internal events CDF or LERF.

5-05 QU-D2 Finding 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)

Some of the significant cutsets are considered to be unrealistic due to the conservatisms related to not modeling alternate mitigating capabilities as discussed in assessment of SR AS-A5.

Recommend reviewing cutsets for modeling consistency after addressing F&O 5-21 relative to SR AS-A5.

The Fire PRA has looked extensively at top cutsets to reduce overall CDF and LERF. This F&O does not impact the Fire PRA.

The internal events PRA was deemed to be conservative without modeling several operator recovery actions. As discussed in F&O 5-21, the value of the additional recovery actions is limited.

Closure of this F&O is not expected to have significant impact on CDF or LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 18 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 5-07 QU-D1 Finding See F&O 5-01(cutsets 7 and 8 are the same except for the flag events).

(This F&O originated from SR QU-D1)

Additional cutset are being generated in the results due to flag events remaining in the model that are not set to TRUE or FALSE, see cutsets 7 and 8 in the final CDF results where the only difference is a flag that is used for some sort of identification.

Recommend setting flag events to either TRUE or FALSE.

The SR is met at CC-III.

This issue was resolved via PRA model update after the Peer Review. The model update is included in the Fire PRA. This finding has no impact on the Fire PRA.

5-08 QU-D5 Finding CN-RAM-14-036, Appendix C describes how the non-significant cutsets from the Rev. 6d model were reviewed. However, this review should be repeated using current model results.

(This F&O originated from SR QU-D5)

Insights from previous model are not indicative of current model, non-significant cutsets review should be performed for each version of the model during quantification updates.

Recommend repeating the review of the non-significant cutsets using the Rev. 7 model results.

This is a documentation issue for the internal events PRA. The non-significant cutsets of Model 6d were representative of the Model 7a under Peer Review. Closure of the F&O is not expected to significantly change the Internal Events PRA model There is no impact on the Fire PRA. Non-significant cutsets reviews were performed for Fire PRA.

5-10 QU-D7 Finding Importance measures should be reviewed to ensure that they make logical sense. Currently, the items are only ordered alphabetically, and there is no accompanying discussions to document the review of the importance measure, either individually or collectively.

(This F&O originated from SR QU-D7)

They could be ordered by FV or RAW to make it easier to compare. Appendix D is titled component importance measures but basic events are displayed, suggest re-title or develop actual component based performances that include all applicable failure modes (e.g., a normally closed valve needing to open or remain closed, depending on the type of sequence).

Recommend performing the review in order of importance to make sure the item importance make logical sense.

This is a documentation issue for the internal events PRA. The list was sorted alphabetically instead of by FV or RAW. This made the review more of a challenge for the Peer Team. No impact on the Fire PRA.

5-11 QU-C2 Finding 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 3-Should be able to run for a while without cooling 4-occur much later 5-occurs somewhat further out in time 6-should be sufficient time 7-should come quickly 8-assumed to be separated in time 9-Assumed to be negligible because these are performed outside of the CR alone 10-Assumed to have different dependence for some initiators vs.

others without sufficient basis 11-Assumed to be independent based on different procedure alone 12-Assumed independent based on different cue alone (This F&O originated from SR QU-C2)

Methodology as performed can skew the results.

Use the HRA Calculator to assess all combinations and systematically evaluate aspects of dependency.

This issue was addressed in the HRA analysis for Fire PRA.

An evaluation was completed using the EPRI HRA calculator for the dependency analysis to conservatively estimate the impact on the internal events PRA. The HRA calculator was used with existing timing with no credit for factors such as intervening success. The CDF increased by approximately The CDF increased by approximately 1.9E-6 per year and LERF by approximately 3.8E-8 per year. These increases are within margin for RG 1.174

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 19 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 5-13 QU-A1 Finding HRA values are not seeded prior to quantification.

(This F&O originated from SR QU-A1)

Combinations that should be above the truncation limit may not be included in the results as the individual HFE values would knock them below the truncation limit.

Recommend utilizing the HRA helper tool to develop the seeding values for HFE events.

This SR Met at CC-III.

This issue was addressed in the HRA analysis for Fire PRA.

An evaluation was completed using the EPRI HRA calculator for the dependency analysis to conservatively estimate the impact on the internal events PRA. The HRA calculator was used with existing timing with no credit for factors such as intervening success. Seeding was done with the EPRI HRA calculator. The CDF increased by approximately 1.9E-6 per year and LERF by approximately 3.8E-8 per year. These increases are within margin for RG 1.174.

5-16 QU-E3 Finding The common cause events and those previously identified in DA-A3 were found not to have uncertainty parameters assessed.

(This F&O originated from SR QU-E3)

May be under estimating the range of uncertainty associated with total CDF.

Define common type codes for similar components, expand the CCF to provide uncertainty metrics and complete the values for type codes that currently are set to zero. Once completed, perform the UNCERT evaluation.

This SR Met at CC-II.

This issue was resolved in a PRA model update following the Peer Review. No impact on the Fire PRA.

5-18 QU-F2 Finding Asymmetry analysis was not performed in the quantification analysis.

(This F&O originated from SR QU-F2)

Insights from alternate alignments may not be adequately categorized or identified.

Recommend performing alternate alignment runs to identify uncertainties and asymmetries.

This SR met at CC-III.

The asymmetry was the only concern since it was not quantitatively assessed. Asymmetry has been qualitatively evaluated and known modeling attributes assessed. Closure of this F&O is not expected to significantly impact CDF and LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 20 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 5-21 AS-A5 Finding Several conservatisms were noted during review of the accident sequences where procedural pathways provide additional potential mitigating capabilities that could be credited and have been credited at other similar plants. As the exclusion of these mitigating pathways has a potential to impact the results of the model and importance measures, these sequences should be reviewed and the additional mitigating capabilities should be incorporated or appropriately dispositioned:

1) Loss of EFW - Given a loss of EFW there is a potential to restore MFW or condensate prior to initiating feed and bleed.

2) HPI Requirement on SSB - The accident sequence and event tree currently require HPI in the event of an SSB regardless of if a bleed feed requirement arises. If the RCS pressure boundary is not breached HPI should not be required as RCS volume should remain sufficient such that core uncover does not occur.

3) RWST in Small LOCAs - Currently the capability to re-fill the RWST and remain on injection given a failure of recirculation is not modeled. RWST re-fill is currently only credited on a SGTR or an ISLOCA.

4) SLOCA where HPI fails - There is no credit for depressurization and low pressure injection currently in the accident sequences. Response was provided by VCS stating that depressurization would not occur before depletion of the CST but does not discuss the capability of re-filling the CST or providing additional source of water from the SSW supply valves, additionally timing to de-pressurize seemed to be long.

(What is the shutoff head of the pumps, it appears that the pressure drops below 600 psi after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).

5) Failure of Recirculation Auto-Swap: Review of cutsets revealed that the failure of relays (due to CCF) in the RHR recirculation valves fails the recirculation function however it does not appear as if an operator action to manually align for recirculation.

• SSB with HFE to stop SI: Based review of the cutsets, failure to stop the SI when not required results in the opening of the PORVs (or SRVs?, unclear based on initial review) however no credit is taken to shut the PORV block valve (block valve should be able to close successfully for RCS pressure regardless of fluid type, steam or water (solid)) or for the SRV to reclose after SI is terminated (after the valves have been challenged, WCAP-11677 discussed how SRVs have been to re-close successfully up to 3 times after passing water).

Additionally, if recirculation did fail and the additional credit for SI termination (after the valves have been challenged) failed, re-fill of the RWST and continued injection may be possible.

(This F&O originated from SR AS-A5)

Exclusion of these alternate mitigating capabilities can skew the results and mask importance.

Recommend dispositioning or incorporating these actions to the extent possible.

The Fire PRA is not impacted. The Fire PRA cutsets have been reviewed extensively to reduce CDF and LERF.

F&O AS-A5 questioned the PRA success criteria for secondary side breaks and small LOCA and lack of credit for additional systems and recovery actions.

The success criteria was questioned for small LOCA and secondary breaks. Success criteria that allow more recovery credit would reduce the conditional core damage probability and large early release probability associated with these initiating events.

Suggested recoveries were crediting main feedwater, manually recovering failure of auto swap to recirculation, closing a PORV block valve for a stuck open PORV and re-filling the RWST for a small LOCA.

Closing the block valve for a stuck open PORV is already in the VCSNS PRA model.

Manual recirculation swap is of almost no benefit in the PRA because it relies on much of the same equipment and instrumentation as the much more reliable auto swap.

Main feedwater is only useful for a limited number of scenarios where offsite power has not been lost, but all three emergency feedwater pumps and feed and bleed have failed. So, the possible credit is very limited.

Credit for re-filling the RWST following a small LOCA is of little numerical value based on:

Sump recirculation must have failed to use this.

The cue to refill occurs when the recirculation swap occurs (18% RWST level) and this limits the time window for success.

The possible re-fill rate of the RWST from Reactor Make-Up (120gpm) will not keep up with small LOCA flow.

Make-up to the RWST from the spent fuel system is another option. Gravity drain from this source would also be unlikely to keep up with small LOCA flowrates.

Based on the reasons above, credit for suggested recoveries would be minimal.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 21 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 5-22 AS-A6 Finding Event trees and discussion for the SSB (in containment, Sec.

7.9) and SLOCAs (Sec. 7.10) include RBCUs to succeed, but these are not required for MLOCA (7.8) and LLOCA (Sec 7.5). It is unclear why they are required for prevention of core damage for these event, for MLOCA and LLOCA the document states they are only needed for containment pressure control and integrity so these should also apply to at least the SLOCAs and the SSBs (in containment).

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

For the SLOCA event tree regardless of success or failure of the RBC or SP function success can be achieved by FB and recirculation. Also for CSLO_RCP (may be others). CSLO (LSP) Event Tree has a similar issue. RCBU and SP failure show leading to core damage if the RCS is depressurized (via SGP), but if SGP fails long term cooling can be successful on recirculation. This seems strange given in one case the RCS pressure has lowered and the other RCS pressure remains high but containment cooling is not questioned but in both cases recirculation is successful. It appears that this should be removed for CD and put into level II.

Recommend determining and clearly document for all related scenarios if RBCU and SP is required to prevent Core Damage or if it is only required for LERF/Level II.

This SR is met at CC-I-III.

The SLOCA Success Criteria has multiple paths.

(1) With one RBCU running, the containment pressure remains below containment spray actuation. Without containment spray the RWST inventory is adequate to provide time to allow operators to reduce RCS pressure. Long term cooling is provided by low head recirculation.

(2) If an RBCU is not running, then containment pressure increases and containment spray is actuated. This reduces available time to depressurize the RCS, so long term high pressure recirculation must be established. In this mode the charging pumps take suction on the RHR pump discharge. The RHR pumps take suction on the containment sumps.

An RBCU is one success path for SLOCA and is included in the model.

For Medium and Large Break LOCA, spray actuation occurs regardless of RBCU operation, so the logic is not necessary.

The model is correct no changes are required to address this finding. This finding has no impact on the Fire PRA.

5-27 AS-B1 Finding Mitigating system dependencies as a result of an initiating events are not clearly described and as modeled will result in skewing of the importance.

(This F&O originated from SR AS-B1)

Impacts of MSLBs to secondary heat removal (loss of a SG) are not clearly stated, isolation of the SG should be considered successful such that the SG is lost for the purposes of heat removal. As currently modeled it assumes that SG isolation will fail.

Recommend more clearly documenting the impacts of initiators on mitigating systems and apply the failure rate equally across the potentially affected equipment (equal failure probability distribution across RCS loops A, B, and C (IEF/3).

The F&O is concerned with documentation of the impact of initiating events on mitigating systems.

The Accident Sequence Notebook is References in the SR road map. The notebook discusses mitigating system dependencies such as charging pump dependencies on CCW. However, these items are not specifically called out or summarized as mitigating system dependencies. Revising documentation of mitigating dependencies to close this F&O will not impact CDF or LERF.

5-28 AS-B2 Finding PRV-Challenge: The basis for this static probability of a pressure challenge is unclear.

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

Aside from an ATWS events, the only pressure challenges should be from a loss of all AFW/MFW.

This could be modeled instead of using the failure probability, failure of AFW could represent the pressure challenge.

Recommend removing static failure probability and model a pressure challenge instead of a probability.

This SR is met at CC-III.

Pressure challenges for the PORVs and Safety valves are included in the fault tree. The Peer Reviewer did not see this gate. Closure of this F&O is not expected to impact CDF or LERF.

6-04 IFSO-A4 Finding The model identifies spray, flood and HELB considerations for piping considerations. Human induced and some other scenarios such as expansion joints were not considered.

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

The provided documentation does not identify the any human induced or expansion joint modeling.

There is qualitative screening of these events but it is unclear if the approach is sufficient. The EPRI evaluation method does contain induced events and expansion joint failure was a significant industry event.

Recommend expanding the model to address potential maintenance induced failures and expansion joint breaches.

This SR is met at CC-III.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

EPRI guidelines for internal flooding probabilistic risk assessment (EPRI 1019194 section 5.6) allow qualitative screening if two or more isolation valves are used. VCSNS Station administrative procedure SAP-0201 states that two valve protection is required for systems or tanks that can cause major flooding of buildings or systems being worked.

Based on this, qualitative screening of human induced flooding was appropriately screened in the VCSNS internal flooding analysis.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 22 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 6-05 IFEV-A1 Finding After a review of the characterization of the scenarios, it was found that the grouping was based on a zone and system but did not consider timing, component impacts based on different line flow rates which may lead to conservative results for risk-significant scenarios.

(This F&O originated from SR IFEV-A1)

Grouping any size pipe, assuming guaranteed failures, and not differentiating impacts can result in over estimation and non-realistic results.

A conservative screening analysis might be acceptable for very low frequency contributors. However, flooding is a major contributor to the risk profile and it is recommended that a more detailed assessment based on individual flooding scenarios be performed.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

The internal events modeling is conservative.

Refinements were not completed at the time of the Peer Review. These refinements will reduce CDF and LERF for internal events and are being tracked under the corrective action program 6-06 IFEV-A6 Finding The initiating event frequency was based on generic information.

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

The data was taken without consideration of plant experience.

Recommend updating the EPRI data using a Bayesian approach or something similar.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

At the time of the internal flood PRA effort there were no plant flood events. The generic information was appropriate for this first application. A Bayesian update will decrease internal flood IE frequencies. These refinements will reduce CDF and LERF for internal events and are being tracked under the corrective action program 6-07 IFQU-A1 Finding No review could be found in Flooding documentation to confirm accident sequence applicability to the flood scenario.

(This F&O originated from SR IFQU-A1)

No review could be found in Flooding documentation to confirm accident sequence applicability to the flood scenario. A formal review is necessary to identify flood scenarios and link them to accident sequences.

A formal review is necessary to identify flood scenarios and link them to accident sequences.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

The internal events modeling is conservative.

Refinements were not completed at the time of the Peer Review. Refinements will reduce CDF and LERF for internal events and are being tracked under the corrective action program.

6-09 IFSN-A3 Finding No documentation of automatic features and operator responses could be found that could terminate or contain flood propagation.

For example, the top cutset which is a BD line break in IB, does not have a failure of automatic isolation of the BD line.

(This F&O originated from SR IFSN-A3)

No documentation of automatic features and operator responses could be found that could terminate or contain flood propagation. For example, the top cutset which is a BD line break in IB, does not have a failure of automatic isolation of the BD line.

For each flood area and flood source, recommend reviewing availability and document any automatic features and operator responses that could terminate or contain flood propagation.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

The internal events modeling is conservative.

Refinements were not completed at the time of the Peer Review. Inclusion of the automatic features will reduce CDF and LERF for the internal events and is being track under the corrective action program.

6-10 IFSN-A4 Finding The capacity of the drains is estimated in Appendix A of notebook CN-RAM-13-046. However, no calculation could be found on water retention in sumps, curbs, dikes, etc.

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

The capacity of the drains is estimated in Appendix A of notebook CN-RAM-13-046. However, no calculation could be found on water retention in sumps, curbs, dikes, etc.

Recommend estimating water retention in sumps, curbs, dikes, etc. and apply the impact on timing calculations on a scenario basis.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

For the flood model, retention at sumps and curbs will increase time for recovery actions. These refinements will reduce CDF and LERF for internal events and are being tracked under the corrective action program.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 23 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 6-12 IFSN-A6 Finding Susceptibility of PRA equipment (by type only) to flood is documented in Section 7.6, Table 7.6-1. However, failures such as jet impingement or pipe whip, for example, are every flood area and source specific, and therefore, first, a more detailed review has to be performed to identify PRA equipment in each flood area, and then identify susceptibility to failure mechanisms.

Also, after flood walkdown, it became clear that steam driven EFW pump is also susceptible to failure due to high humidity in the IB due to HELB since the room that contains the pump has a vent opening big enough to impact conditions inside that room.

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

Susceptibility of PRA equipment (by type only) to flood is documented in Section 7.6, Table 7.6-1.

However, failures such as jet impingement or pipe whip, for example, are very flood area and source specific, and therefore, first, a more detailed review has to be performed to identify PRA equipment in each flood area, and then identify susceptibility to failure mechanisms. Also, after flood walkdown, it became clear that steam driven EFW pump is also susceptible to failure due to high humidity in the IB due to HELB since the room that contains the pump has a vent opening big enough to impact conditions inside that room.

Refer to resolution of IFSN-A5 and consider modeling failure of the TD-EFW pump due to HELB in IB or justify its survivability.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

The flood model assumed that electric equipment within the room with the flood source would fail.

There were no specific analysis to eliminate equipment, except in the case where the equipment was specifically designed to be wetted.

The documentation of the excluded equipment is provided throughout the calculations and associated databases/spreadsheets. These were was not listed in a specific table. Documentation could be updated to facilitate Peer Review, but overall modeling of flood remains conservative.

The HELB impact on the TD-EFW pumps is covered in the EQ program. The TD-EFW pump is a mechanical device with no specific components on the EQ list.

The pump is qualified to operate in accident conditions.

Further, a detailed GOTHIC model of HELB in question was completed for a separate project.

The HELB pressure pulse in the noted IB location is less than 0.1 psi and decreases to atmospheric.

There is limited motive force to change the environment through the vent area on top of the TD-EFW pump room.

6-13 IFSN-A8 Finding PRA notebook CN-RAM-13-044 Section 7.4 states that drains and backflow have been considered in the flooding analysis.

However, no evidence could be found on actually considering drains and backflow for inter-area flood propagation.

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

PRA notebook CN-RAM-13-044 Section 7.4 states that drains and backflow have been considered in the flooding analysis. However, no evidence could be found on actually considering drains and backflow for inter-area flood propagation.

Recommend using flood propagation through drains and backflow.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

Back flow has been considered in the flood model.

The peer review team gave a Best Practice for the VCSNS flood propagation calculation. It covers how flood water flows down to lower elevations and may propagate into other areas. Equipment hatches, stairwells and drains are included. Note that F&O 6-10 specifically calls out that drain flow was calculated. This F&O is resolved.

6-14 IFSN-A9 Finding Flood rate calculations are performed using EPRI Tech Report formula, and compared with the system flow rates, and the lower is used in the timing calculation. This may be non-conservative since in some cases lower than expected flow rates may be used. Suggestion: Timing calculations are performed based on estimation of flood height in an 8 hr time period, then the available time before PRA equipment fails in the area using the break flow rate. It is suggested to calculate the timing based on first estimating the critical flood volume and then dividing it by the break flow rate.

(This F&O originated from SR IFSN-A9)

Flood rate calculations are performed using EPRI Tech Report formula, and compared with the system flow rates, and the lower is used in the timing calculation. This may be non-conservative since in some cases lower than expected flow rates may be used. Suggestion: Timing calculations are performed based on estimation of flood height in an 8 hr time period, then back calculating the available time before PRA equipment fails in the area using the break flow rate. It is suggested to calculate the timing based on first estimating the critical flood volume and then dividing it by the break flow rate.

Recommend using pump runout flow rates instead of normal system flow rates, for each system, and compare with the break flow rate calculated with the EPRI formula.

Consider calculating the timing based on the methods provided in the F&O.

This SR is met at CC-III The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

The modeling of critical volumes and runout flows where appropriate will produce more representative times for assessing flood impacts and operator recovery times. Refinements to the flooding PRA had not been completed at the time of the Peer Review. The refinements are expected to result in in lower CDF and LERF for internal events and are being tracked under the corrective action program

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 24 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 6-15 IFQU-A6 Finding 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 increase of five minutes was added to Texe and Tcog." This appears to have no technical basis.

(This F&O originated from SR 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 increase of five minutes was added to Texe and Tcog."" This appears to have no technical basis.

Recommend interviewing operators on the impact flooding on their actions. Consider impact of flooding on HEPs only if the flood creates additional stress, causes access restrictions, or if the operations lack training on response to flooding events. Consider impact only if the operator action is expected to be performed before the flood is isolated/over.

The SR is met at CC-III.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

It is standard practice to adjust external events PRA timing to account for more stress. Changes that were made in CN-RAM-13-048 are expected to be conservative. No significant changes to CDF and LERF are expected to occur during resolution of this F&O.

6-17 SY-A4 Finding No system design walkdowns were found to support recent modeling updates.

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

The SR requires that a plant walkdown and interviews with plant staff be performed to ensure the PRA models the as-built, as-operated condition.

Recommend performing and documenting a systems walkdown. It is suggested that the walkdown team include operations staff and/or maintenance staff.

Walkdowns for the IPE (original internal events PRA analysis) are document in DC00300-019, the Plant Walkdown Notebook.

Walkdowns for the Fire PRA are described in DC00340-001 Task 5.16, Fire PRA Plant Final Report.

Walkdowns for PRA applications have been conducted for Fire, Seismic and Flooding between 2014 and present. All features that affect the PRA model for this application were included. Closeout of this F&O is not expected to impact CDDF or LERF.

6-18 SY-A11 Finding The system models exclude in many cases component failures on the basis of guidance in PSA-01. For example; Locked open manual valves XVG-8471A, B, C and normally open manual valve XVG-8388 are not modeled. No quantitative considerations are specified to meet SY-A15.

(This F&O originated from SR SY-A11)

The screening criteria for component failures as defined by SY-A15 is quantitative. The current screening criterion is qualitative based on generic impacts which cannot determine actual failure mode importance. For time-dependent failures it is important to consider exposure time. The qualitative approach has not provisions for this aspect of the failure rate.

Recommend defining a single process for quantitative screening based on the component at a system/train level and reapply. Incorporate restored events to the model.

A sensitivity study has been completed to assess the impact of the quantitative screening criteria.

The increase in CDF and LERF was less than 0.3%. As the screening criteria is applied to the formal PRA model no significant changes are expected for the internal events or Fire PRA.

6-19 SY-A13 Finding 1/3 rule used and too many divergence paths not addressed.

(This F&O originated from SR SY-A13)

The use of the 1/3 rule as defined in the plant PRA guidelines is no longer utilized since this neglects the effects of even small leakage on closed loops systems with finite volume. The divergence screening should also meet the quantitative requirements of SY-A15 but no quantitative basis is provided.

Recommend defining a single process for assessing divergence path and highlight consideration such as design margin, leakage isolation (auto or manual), and system capacity.

A sensitivity study has been completed to assess the impact of the quantitative screening criteria.

The increase in CDF and LERF was less than 0.3%. As the screening criteria is applied to the formal PRA model no significant changes are expected for the internal events or Fire PRA.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 25 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 6-20 SY-A14 Finding Overall modeling is considered consistent with the data notebook and typical generic databases. There are a few instances where the models are inconsistently applied across some of the notebooks due to the use of screening. Some other failure modes were excluded as failures although data existed in the data notebook and other generic sources. An example is the exclusion of air dryers as a failure mode. A suggestion is made that the models be reviewed for compliance to the standard and that all modeling assumptions be based on a single standard. Currently there are two Referenced standards that may be in conflict (PSA-01 and CN-RAM-13-020). An example is the statement in the CC notebook:

""Modeling note 18 below was obtained from Reference 22.

Note that some modeling notes may require additional documentation or changes due to differences with the screening criteria developed in Reference 13.

(This F&O originated from SR SY-A14)

Overall modeling is considered consistent with the data notebook and typical generic databases.

There are a few instances where the models are inconsistently applied across some of the notebooks due to the use of screening. Some other failure modes were excluded as failures although data existed in the data notebook and other generic sources. An example is the exclusion of air dryers as a failure mode. A suggestion is made that the models be reviewed for compliance to the standard and that all modeling assumptions be based on a single standard. Currently there are two Referenced standards that may be in conflict (PSA-01 and CN-RAM-13-020). An example is the statement in the CC notebook: "Modeling note 18 below was obtained from Reference 22. Note that some modeling notes may require additional documentation or changes due to differences with the screening criteria developed in Reference 13."

Define a single process for quantitative screening based on the component at a system/train level and reapply.

Incorporate restored events to the model.

The vendor had been tasked with updating the screening procedure to current industry standards.

The Peer Review team took the existing procedure and the new recommendations as a delta in screen applications.

A sensitivity study has been completed to assess the impact of the quantitative screening criteria.

The increase in CDF and LERF was less than 0.3%. As the screening criteria is applied to the formal PRA model no significant changes are expected for the internal events or Fire PRA.

6-21 SY-A15 Finding It appears that the original system fault tree screening was developed, or at least consolidated in PSA 01, (Attachments I/II) but was later updated in CN-RAM-13-020 as an assessment revealed that the PSA-01 screening criteria were insufficient to meet the standard. CN-RAM-13-020 also suggests that PSA-01 be updated with the new guidance. However, the CCW system NB (CN-RAM-14-022) lists both of these documents as References (References 13 and 21) while in contrast the EF Notebook only lists the updated document CN-RAM-13-020.

Section 7.3.1 of the CCW NB says on page 36 that some modeling notes may require additional documentation or changes due to differences with the screening criteria developed. It appears that the revised guidance was not utilized in a consistent assessment approach.

(This F&O originated from SR SY-A15)

Screening appears to be performed using two different documents that in some cases are contradictory. It appears that the original system fault tree screening was developed, or at least consolidated in PSA 01, (Attachments I/II) but was later updated in CN-RAM-13-020 as an assessment revealed that the PSA-01 screening criteria were insufficient to meet the standard.

Recommend defining a single process for quantitative screening based on the component at a system/train level and reapply. Incorporate restored events to the model.

A sensitivity study has been completed to assess the impact of the quantitative screening criteria.

The increase in CDF and LERF was less than 0.3%. As the screening criteria is applied to the formal PRA model no significant changes are expected for the internal events or Fire PRA.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 26 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 6-22 SY-A18 Finding The internal flooding model indicates that isolated flooding events will be accounted for in the internal events modeling.

However, no contribution was defined in the initiating event frequency of the internal events. Therefore the isolation of CCW or SW given a successful flood isolation (state FS02 in the IF notebook). Is not present in the model to cause an isolation of the system response model.

The documentation does not sufficiently support successful operation of the CSIP or RHR pumps given a lack of room cooling. The analysis from the IPE includes door opening to demonstrate marginal performance for the CSIP room when no HVAC is present. No HVAC is modeled or discussed in the CSIP notebook and no compensatory HFE was found in the HRA notebook.

(This F&O originated from SR SY-A18)

The internal flooding model indicates that isolated flooding events will be accounted for in the internal events modeling. However, no contribution was defined in the initiating event frequency of the internal events. Therefore the isolation of CCW or SW given a successful flood isolation (state FS02 in the IF notebook). Is not present in the model to cause an isolation of the system response model.

The documentation does not sufficiently support successful operation of the CSIP or RHR pumps given a lack of room cooling. The analysis from the IPE includes door opening to demonstrate marginal performance for the CSIP room when no HVAC is present. No HVAC is modeled or discussed in the CSIP notebook and no compensatory HFE was found in the HRA notebook. The omission of an initiating event of higher frequency (more likely to be isolated than not) could impact total frequency. HVAC has gained importance at sites when examined in more detail and the omission appears to leave out common mode failure for the ESFAS.

Include the isolated floods in the internal events frequency. Review the HVAC assessment to define a clear success criteria for room cooling requirements.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

The flooding model was a conservative analysis that assumed loss of the first PRA modeled component led to core damage. As refinements are made to flooding model to address the conservatisms and associated F&Os, the CDF and LERF are expected to improve.

See F&O 6-23 for the question on HVAC.

6-23 SY-A22 Finding The supporting HVAC information from notebook documentation (CN-COA-91-129) indicates that "Among other things, the HVAC system provides cooling of the RHR/Spray Pump room and the Charging/SI pump room. The other rooms inside the intermediate building have been eliminated from consideration. An analysis was performed to determine if a failure of HVAC would cause a subsequent failure of the pumps located within these 2 rooms. The analysis results, presented in Reference 8, indicate that HVAC is not required as long as the doors are open to provide natural circulation cooling in these pump rooms."

This appears to indicate that some form of cooling is required for the CSIP and RHR/Spray pumps. No HVAC cooling requirement is documented in the ECCS notebook and not operator actions are defined to open doors for these rooms.

(This F&O originated from SR SY-A22)

The supporting HVAC information from notebook documentation (CN-COA-91-129) indicates that "Among other things, the HVAC system provides cooling of the RHR/Spray Pump room and the Charging/SI pump room. The other rooms inside the intermediate building have been eliminated from consideration. An analysis was performed to determine if a failure of HVAC would cause a subsequent failure of the pumps located within these 2 rooms. The analysis results, presented in Reference 8, indicate that HVAC is not required as long as the doors are open to provide natural circulation cooling in these pump rooms." This appears to indicate that some form of cooling is required for the CSIP and RHR/Spray pumps. No HVAC cooling requirement is documented in the ECCS notebook and not operator actions are defined to open doors for these rooms. Therefore the model potentially is utilizing the ESFAS pumps outside their operating range.

Review the HVAC assessment to define a clear success criteria for room cooling requirements.

The SR map for the Peer Review team was not adequate and did not provide the appropriate documentation. The Reference 8 is the room heat up calculation for the pump rooms. The calculation demonstrates that doors do not have to be open for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time. The calculation was not provided to the Peer Review team.

A separate calculation was completed for the electrical room HVAC. That calculation showed that the electrical room doors needed to be opened at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The model includes AADRXSW1DAHE and ABDRXSW1DBHE.

These model the failure to open the doors to the electrical rooms. The electrical room calculation was not provided to the Peer Review team.

The issue is considered resolved from the technical perspective. Documents need to be updated for consistency and proper Reference.

Closure of this F&O is not expected to impact CDF or LERF.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 27 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 6-24 SY-B6 Finding Review of the system notebooks indicates a thorough analysis of each system that identifies support systems and varying operating conditions which may impact accident scenarios.

HVAC not addressed for CSIP and RHR. There are no analyses or discussions provided for other rooms.

(This F&O originated from SR SY-B6)

The system notebooks do not appear to handle HVAC in a complete manner. The discrepancy is discussed in SY-A22 for two areas but there is no clear guidance as to why other rooms do not require cooling.

Review the HVAC assessment to define a clear success criteria for room cooling requirements.

The SR map for the Peer Review team was not adequate and did not provide the appropriate documentation. The pump room heat up calculation demonstrates that doors do not have to be open for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time.

A separate calculation was completed for the electrical room HVAC. That calculation showed that the electrical room doors needed to be opened at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The model includes AADRXSW1DAHE and ABDRXSW1DBHE.

These model the failure to open the doors to the electrical rooms. The electrical room calculation was not provided to the Peer Review team.

The issue is considered resolved from the technical perspective. Documents need to be updated for consistency and proper Reference.

Closure of this F&O is not expected to impact CDF or LERF 6-25 SY-B12 Finding HVAC has been screened for pump rooms in some cases based on the expected conditions in the room with the door open.

Given that the doors are normally closed and that no operator action is modeled the implication is that they are essentially taking credit for an unmolded action.

(This F&O originated from SR SY-B12)

HVAC has been screened for pump rooms in some cases based on the expected conditions in the room with the door open. Given that the doors are normally closed and that no operator action is modeled the implication is that they are essentially taking credit for an unmolded action.

If resolution of finding for SY-A22 indicates that cooling is required, recommend model HVAC to the rooms and/or operator actions to compensate if unavailable based on current plant operating procedures and practice.

The SR map for the Peer Review team was not adequate and did not provide the appropriate documentation. The pump room heat up calculation demonstrates that doors do not have to be open for the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time.

A separate calculation was completed for the electrical room HVAC. That calculation showed that the electrical room doors needed to be opened at 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The model includes AADRXSW1DAHE and ABDRXSW1DBHE.

These model the failure to open the doors to the electrical rooms. The electrical room calculation was not provided to the Peer Review team.

The issue is considered resolved from the technical perspective. Documents need to be updated for consistency and proper Reference.

Closure of this F&O is not expected to impact CDF or LERF 6-26 SY-B13 Finding This should be coupled with the issues with SY-A15(a). In addition to showing the screened components do not impact the system there should also be an assessment that they do not impact multiple systems (i.e. many of the support systems that use screening).

(This F&O originated from SR SY-B13)

This SR indicates that screening for support systems (require for operation of other systems) should be screened based not only on the support system impact but also when considering quantitative impact on those system(s) being supported. Since no quantitative screening was developed and no assessment of pass through contributions some components could be excluded that should be included.

Link to SY-A15. Also expand screening to address supporting system impact on frontline systems The Dependency Notebook contains the Support System Matrix. No support system dependencies were screened from consideration. The Dependency Notebook was not listed in the Supporting Requirement road map and was not supplied to the Peer Review team. This F&O is considered resolved and there is no impact on the Fire PRA.

Document Control Desk - Attachment 8 LAR-16-01490 RC-18-0091 Page 28 of 28 ID SR Type Description Basis for Significance Possible Resolution Fire PRA Impact 6-27 SY-B14 Finding There are several canned statements in all of the notebooks.

For example:

1) The systems and components credited in the PRA model are expected to function in the environments anticipated following events and these ensuing environments are not expected to exceed component qualifications.

2) The environmental conditions are expected to be similar to those expected for design basis events.

3) No environmental conditions are expected that will exceed the operating qualifications of the components.

4) No non-qualified equipment is credited for operation in harsh environments.

5) The impact of high energy line breaks is addressed in the internal flooding analysis.

Systems and components are credited for conditions and loads expected during the initiating events modeled.

(This F&O originated from SR SY-B14)

The notebooks provide a generic statement indicating that the components are environmentally qualified without a corresponding Reference to an analysis for operation in environments that may exceed design basis. The analysis is needed to support review, update and applications.

Recommend expanding the generic statement to include specific qualification assessments, component qualification ranges, and expected environmental challenges such as high temperature, HELB, humidity, radiation, etc.

This SR is met at CC-I-III.

The system notebooks did not Reference the appropriate EQ information sources for equipment.

Equipment (e.g. EFW pumps, CCW pumps) are qualified for harsh environment. Closure of this F&O is not expected to affect CDF and LERF for Internal Events or Fire PRA.

6-28 IFEV-A7 Finding The documentation performed a qualitative screening and did not estimate any contribution due to HIF. This is inconsistent with the EPRI guidance documented used to develop the internal flooding assessment.

(This F&O originated from SR 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.

Expand the model to address potential maintenance induced failures and expansion joint breaches.

The Fire PRA does not model flooding as an initiating event. The Fire PRA is not impacted.

In addition to the limited maintenance activities, EPRI guidelines for internal flooding probabilistic risk assessment (EPRI 1019194 section 5.6) allow qualitative screening if two or more isolation valves are used. Station administrative procedure SAP-0201 states that two valve protection is required for systems or tanks that can cause major flooding of buildings or systems being worked. Based on this, qualitative screening of (HIF) human induced flooding was appropriate in the VCSNS internal flooding analysis. Resolution of this F&O is not expected to significantly impact internal events CDF or LERF.

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 1 of 9 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LAR-16-01490 ENCLOSURE 1 ATTACHMENT 9 NFPA 805 Chapter 3 Requirements for Approval 10 CFR 50.48(c)(2)(vii)

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 2 of 9 1.0

SUMMARY

DESCRIPTION In addition to the request for approval of the changes to the modification scope, this LAR also includes requests for approval of the following:

Request for Approval of Performance-Based Alternatives for Chapter 3 NFPA 805 (10 CFR 50.48(c)(2)(vii))

o NFPA 805 Section: 3.3.5.1 Wiring Above Suspended Ceiling This is a revision to the original approval request L2. A revised approval request is attached.

o NFPA 805 Section: 3.3.4 Insulation Materials This is a new approval request L14.

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 3 of 9

2.0 INTRODUCTION

FOR APPROVAL REQUEST L2 - NFPA 805 SECTION: 3.3.5.1 WIRING ABOVE SUSPENDED CEILING NRC approval of Approval Request L2 was granted in the VCSNS Safety Evaluation, dated February 11, 2015, Section 3.1.4.2 (Reference 1). During third party reviews additional areas in the plant where wiring is installed above suspended ceilings were identified in the plant.

Specific revisions to the NRC acceptance of Approval Request L2 in the VCSNS Safety Evaluation include:

Revising the Request to simplify the statement that there are areas which have the non-compliant wiring. Additionally, SCE&G would like to clarify that the wiring consists of power cables, control cables, and energy limited communications cables, which are not listed for plenum use.

Revising the Basis for Request to include much of the discussion from FPE RAI 08, which was evaluated and approved by the NRC in the Safety Evaluation. The additional information identifies that the predominant type of wiring is for lighting with other installations of communication cables (phone, computer, video and security). These low energy communication cables are not considered a potential source of fire. Also included is the identification of installations with exposed IEEE-383-1974 and non-IEEE-383-1974 rated cable. SCE&G identifies that there are detection and automatic suppression systems installed in most of these areas which mitigates the hazard.

In the Nuclear Safety and Radiological Release Performance Criteria, SCE&G previously stated that engineering specifications and procedures are used to limit the amount of wiring above suspended ceilings. This statement is clarified to state the usage of engineering guidelines and specifications. SCE&G also has identified the presence of detection and/or suppression systems as protection for these areas and that there are no nuclear safety or radiological concerns.

In the Safety Margin and Defense-in-Depth, SCE&G has added additional information on the protection of the wiring above suspended ceilings to support the maintenance of the inherent safety margin. SCE&G also has clarified the types of non-compliance for wiring above suspended ceilings in the defense-in-depth discussion. Clarification was also added to include the protection of structures, systems, and components important to safety.

SCE&G also completed minor editorial corrections throughout.

2.1 Approval Request L2 - NFPA 805 Section: 3.3.5.1 Wiring Above Suspended Ceiling

2.1.1 Request

Approval is requested for existing wiring above suspended ceilings. While the code section is prescriptive in the use and limitation of exposed electrical wire above suspended ceilings, there is existing wiring that does not meet this requirement. This wiring consists of power cables, control cables, and energy limited communications cables that are not listed for plenum use.

2.1.2 Basis for Request:

Station specifications govern the installation of wiring above suspended ceilings. Wiring is specified to be within metal conduits, cable trays, armored cable, or rated for plenum

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 4 of 9 use. The use of suspended ceilings is limited in risk significant areas important to the NSCA, Fire PRA, and NPO analysis. The predominant cable found in these areas is used for lighting and meets the requirements of NFPA 805, Section 3.3.5.1.

Communication cables (phone, computer, video, and security) are not typically routed in conduit. However, this type of cable is low energy and not considered a potential source of fire. Exceptions to the above are an as-found condition with small open trays above the Control Room suspended ceiling that contain IEEE-383-1974 qualified power and control cables for the HVAC equipment located in the overhead space. Another exception is the presence of minor amounts of IEEE-383-1974 qualified and unqualified low voltage power and control cables above suspended ceilings in limited areas. The majority of power production areas with suspended ceilings are designed with detection systems that are located above and/or below the suspended ceilings. Most areas are also provided with an automatic suppression system above and/or below the suspended ceilings. VCSNS complies with the intent of NFPA 805, Section 3.3.5.1 as most cables located above suspended ceilings are directly compliant, and there is a limited installation of non-compliant cables, whose additional hazards are mitigated by the presence of fire seals, detection, and/or automatic suppression systems.

2.2 Acceptance Criteria Evaluation:

2.2.1 Nuclear Safety and Radiological Release Performance Criteria:

The use of limited amounts of wiring above suspended ceilings is restricted by engineering guidelines and specifications with suitable fire protection features present in the area that ensure for the control of combustibles, separation distance, suppression, fire barriers, and protection of the nuclear safety performance criteria as applicable and identified by VCSNS and NFPA 805 Section 1.5, Performance Criteria. The hazard from the limited use of wiring not meeting these requirements is mitigated by the presence of detection and/or suppression systems. There is no nuclear safety or radiological concern from wiring above suspended ceilings.

2.2.2 Safety Margin and Defense-in-Depth:

The installation of wiring above suspended ceilings is limited. Where installed, the wiring is specified to be enclosed within metal conduits, cable trays, armored cables, or plenum rated through fire testing. The installations of cables not meeting these requirements are protected by the presence of detection and/or suppression systems. The wiring is also mostly energy limited communications cables with limited amounts of power and control cables. The controls in place to restrict the installation of wiring above suspended ceilings and to protect the wiring when installed meet the intent of the NFPA Chapter 3 requirement. Therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.

The three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control, and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems, and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed.

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 5 of 9 The installation of wiring above suspended ceilings does not affect element (1) as there are limited areas which contain suspended ceilings, and there are limited installations of cables which are not routed in conduit, metal trays, or rated for plenum use. VCSNS has procedures, specifications, and guidelines which govern the installation of wiring above suspended ceilings in these areas. Elements (2) and (3) are not affected as there are limited installations of cables that are not routed in conduit, metal trays, or rated for plenum use, which does not impact the functions of the detection system, automatic and manual fire suppression activities, or the ability of the barriers to restrict the passage of smoke and flame to protect structures, systems, and components important to safety.

Therefore, the defense-in-depth measures are maintained.

2.2.3 Future Installations:

Current plant guidelines and specifications regarding the installation of cables and wiring above suspended ceilings include limitations to prevent similar non-compliant installations in the future.

2.3

==

Conclusion:==

VCSNS determined that the Fire Protection Program engineering and administrative features and controls provide a level of risk management and performance that achieves the following criteria:

Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; Maintains safety margins; and Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 6 of 9

3.0 INTRODUCTION

FOR APPROVAL REQUEST L14 - NFPA 805 SECTION: 3.3.4 INSULATION MATERIALS A new 10 CFR 50.48(c)(2)(vii) Approval Request L14 is needed for thermal insulation materials. This new request identifies the discovery of thermal insulation materials which do not meet the explicit requirements of NFPA 805, Section 3.3.4. Approval is requested to allow existing and future usage of thermal insulation materials that do not meet the requirements of NFPA 805 Section 3.3.4 at VCSNS.

3.1 Approval Request L14 NFPA 805 Section: 3.3.4 Insulation Materials

3.1.1 Request

Approval is requested to allow existing and future usage of thermal insulation materials that do not meet the requirements of NFPA 805 Section 3.3.4 at VCSNS.

3.1.2 Basis for Request:

NFPA 805 Section 3.3.4 states:

Thermal insulation materials, radiation shielding materials, ventilation duct materials, and soundproofing materials shall be noncombustible or limited combustible.

NFPA 805 Section 1.6.36 has re-defined earlier definitions of noncombustible material to the now current definition of limited combustible material:

Material that, in the form in which it is used, has a potential heat value not exceeding 3500 Btu/lb (8141 kJ/kg) and either has a structural base of noncombustible material with a surfacing not exceeding a thickness of 1/8 in. (3.2 mm) that has a flame spread rating not greater than 50, or has another material having neither a flame spread rating greater than 25 nor evidence of continued progressive combustion, even on surfaces exposed by cutting through the material on any plane.

VCSNS has identified several areas within the Power Block where thermal insulation materials are installed for insulation of system piping and equipment for anti-sweat purposes, as well as for personnel protection (overhead bump hazards). Many of these materials meet the criteria in Branch Technical Position (BTP) APCSB 9.5-1 Appendix A/Appendix R for limited combustibles (flame spread rating less than 25), which is consistent with the previous licensing basis for VCSNS. However, there are some insulation materials (Vimasco material and unmarked material) that do not meet this previous BTP APCSB 9.5-1 Appendix A/Appendix R requirement.

Under the current licensing basis, none of the installed thermal insulation materials meet the more restrictive NFPA 805 requirement for limited combustibles (NFPA 805, Section 1.6.36), as their potential heat value exceeds 3500 Btu/lb.

When used for anti-sweat applications, insulation materials that meet the flame spread rating of less than 25 can be found on refrigerant piping associated with heating, ventilation and air conditioning systems and to encapsulate piping and equipment within the Power Block structures. The Vimasco material is generally used on larger piping.

The use of these same materials for personnel protection (overhead bump hazards)

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 7 of 9 typically involves installations that are small in amount and weight and affixed to steel components or structures. As such, the installations of insulation materials for personnel protection are not considered a significant hazard and are considered to be acceptable.

The basis for the approval request of this deviation is:

The forms in which the thermal insulation is installed, and the anticipated conditions meet the intent of the revised limited combustible material definition where the materials used have a flame spread rating of 25 or less and do not support continued progressive combustion. The testing of materials to determine flame spread is in accordance with ASTM E-84. Where the thermal insulation materials exceed the NFPA 805 heat value of 3500 Btu/lb, they do not contribute significantly to fire due to flame spread ratings of 25 or less per ASTM E-84.

Where thermal insulation materials do not meet the flame spread rating of 25 or less, in-depth fire modeling analysis is completed to assess the impact on the existing fire scenarios.

The forms in which the thermal insulation is installed, and the conditions anticipated do not impact the three echelons of defense-in-depth:

Echelon 1 - to prevent fires from starting (i.e., combustible/hot work controls).

Echelon 1 is not impacted because the thermal insulation does not introduce new ignition sources and presents a negligible hazard in terms of secondary or intervening combustibles. For thermal insulation with a flame spread of 25 or less, the anticipated conditions in which it is installed meets the intent of the revised limited combustible material definition because the materials used do not support continued progressive combustion. Where the materials do not have a flame spread rating of 25 or less, the thermal insulation does not introduce new ignition sources and the Fire PRA assesses the impact of the material as a hazard.

Echelon 2 - rapidly detect, control and extinguish fires that do occur thereby limiting damage (i.e., fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans).

Echelon 2 is not impacted because the limited applications of exposed thermal insulation materials installed for industrial personnel safety and on miscellaneous system piping do not result in increased combustible loading which would challenge the design bases of the installed fire protection systems. The presence of the thermal insulation and engineering change process controls do not impact the ability of the automatic suppression and automatic detection systems to perform credited functions, as the materials are of limited quantities.

Echelon 3 - provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (i.e., fire barriers, fire rated cable, success path remains free of fire damage, recovery actions).

Echelon 3 is not impacted because the limited applications of exposed thermal insulation materials installed for industrial personnel safety and on miscellaneous system piping do not adversely impact the installed fire protection systems and features. Additionally, essential safety functions are

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 8 of 9 maintained and capable of being performed. The insulation material does not impact the post-fire safe shutdown capability in any of the fire areas where it is installed.

The Fire Probabilistic Risk Analysis (Fire PRA) currently includes the combustible insulation (material that meets the BTP APCSB 9.5-1 criteria and material that does not) in the fire modeling methodology. Therefore, the Fire PRA Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) include any potential contribution from the thermal insulation.

The forms in which the thermal insulation is installed, and the conditions anticipated do not impact Nuclear Safety. The limited applications of exposed thermal insulation materials do not compromise post-fire safe shutdown capability as previously designed, revised and analyzed. Adequate defense-in-depth measures are in place as described above to ensure that essential safety functions are maintained and capable of being performed.

3.2 Acceptance Criteria Evaluation:

3.2.1 Nuclear Safety and Radiological Release Performance Criteria:

The use of thermal insulation materials that are other than noncombustible and more than limited combustible in the plant does not adversely impact nuclear safety. Identified installations of these insulation materials were included in the Fire PRA methodology.

Therefore, the impact of the materials on fire risk is accounted for in the Fire PRA. The limited applications of exposed thermal insulation materials are not anticipated to compromise post-fire safe shutdown capability as previously designed, reviewed and considered. Adequate defense-in-depth measures are maintained and evaluated to ensure that essential safety functions are maintained and capable of being performed.

The use of insulation material other than noncombustible and more than limited combustible does not impact the radiological release performance criteria. The radiological release review was performed based on areas containing or potentially containing radioactive materials and is not dependent on the type of thermal insulation material. The insulation material, regardless of heat contribution value, does not change the radiological release evaluation performed that concluded that potentially contaminated water is contained, and smoke is monitored. The insulation materials do not add additional radiological materials to the area or challenge systems boundaries.

3.2.2 Safety Margin and Defense-in-Depth:

Identified installations of thermal insulation materials other than noncombustible and more than limited combustible were included in the Fire PRA methodology. Therefore, the impact of the materials on fire risk is accounted for in the Fire PRA. The selection and application of thermal insulation material is controlled by piping and equipment thermal insulation specifications. Procedures which govern the Engineering Change Process are in place to review future installation impacts to the Fire Protection Program and Fire PRA, resulting in updates to the applicable analyses and calculations as required. The insulation material, and specifically the increase in heat contribution in conjunction with the limited applications meets the intent of the NFPA 805 Chapter 3

Document Control Desk - Attachment 9 LAR-16-01490 RC-18-0091 Page 9 of 9 requirement. Therefore, the safety margin inherent within the NFPA 805 Chapter 3 requirement is maintained.

The three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control, and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems, and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed.

The use of insulation materials other than noncombustible and more than limited combustible does not affect elements (1), (2), or (3). The Fire PRA and procedures governing the future installation of these insulation materials ensures the continued availability and reliability of the fire protection systems and features. As such, there is no impact on VCSNSs ability to prevent fires, the functions of the fire detection and automatic and manual fire suppression activities, or the ability of the barriers to restrict the passage of smoke and flame. Therefore, the defense-in-depth measures are maintained.

3.2.3 Future Installations:

Future installations of these insulation materials are controlled by the Engineering Change Process. This control will ensure that the necessary fire protection reviews are completed, and new installations are evaluated for the potential affects that they could have on the fire safety analyses and the Fire PRA. The analysis completed for this approval request will be maintained and will be updated should future installations require changes.

3.3

==

Conclusion:==

NRC approval is requested for the existing and future use of thermal insulation materials that do not meet the heat value content criteria of NFPA 805 Section 3.3.4. VCSNS has determined that the above specified approach achieves the following criteria:

Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; Maintains safety margins; and Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

Document Control Desk - Attachment 10 LAR-16-01490 RC-18-0091 Page 1 of 4 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LAR-16-01490 ENCLOSURE 1 ATTACHMENT 10 Operating License Condition Markup

Document Control Desk - Attachment 10 LAR-16-01490 RC-18-0091 Page 2 of 4 (18)

Fire Protection Program South Carolina Electric and Gas Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated 11/15/11, (and supplements dated 1/26/12, 10/10/12, 2/1/13, 4/1/13, 10/14/13, 11/26/13, 1/9/14, 2/25/14, 5/2/14, 5/11/14, 8/14/14, 10/9/14, and 12/11/14, and 8/30/2018) and as approved in the safety evaluations dated February 11, 2015 and XXXXX. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

a.

Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant.

Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

1.

Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

2.

Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10-7/year (yr) for CDF and less than 1x10-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

b.

Other Changes that May Be Made Without Prior NRC Approval

1.

Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is

Document Control Desk - Attachment 10 LAR-16-01490 RC-18-0091 Page 3 of 4 functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:

"Fire Alarm and Detection Systems" (Section 3.8);

"Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9);

"Gaseous Fire Suppression Systems" (Section 3.10); and "Passive Fire Protection Features" (Section 3.11).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

2.

Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation dated February 11, 2015, to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

c.

Transition License Conditions

1.

Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. and 3. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.

Document Control Desk - Attachment 10 LAR-16-01490 RC-18-0091 Page 4 of 4

2.

The licensee shall implement the modifications to its facility, as described in Attachment S, Table S-1, "Plant Modifications Committed," of SCE&G letter RC-14-0196, dated December 11, 2014, as modified by Enclosure 1, Attachments 2 and 3, of SCE&G letter RC-18-0091, dated August 29, 2018. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications issuance of the safety evaluation.

3.

The licensee shall implement items listed in Attachment S, Table S-2, "Implementation Items," of SCE&G letter RC-14-0196, dated December 11, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by March 31, 2016 as follows:

a.

Items 3, 6, 7, 8, 10, 13, 14, 17, 19, and 21 within 180 days of NRC approval.

b.

Items 1, 2, 4, 11, and 12 by December 31, 2015.

c.

Items 5, 15, 16, 18, 20, 22, and 23 by March 31, 2016.

Document Control Desk - Attachment 11 LAR-16-01490 RC-18-0091 Page 1 of 4 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LAR-16-01490 ENCLOSURE 1 ATTACHMENT 11 Revised Operating License Condition

Document Control Desk - Attachment 11 LAR-16-01490 RC-18-0091 Page 2 of 4 (18)

Fire Protection Program South Carolina Electric and Gas Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated 11/15/11, (and supplements dated 1/26/12, 10/10/12, 2/1/13, 4/1/13, 10/14/13, 11/26/13, 1/9/14, 2/25/14, 5/2/14, 5/11/14, 8/14/14, 10/9/14, 12/11/14, and 8/30/2018) and as approved in safety evaluations dated February 11, 2015 and

________. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

a.

Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant.

Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

1.

Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

2.

Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10-7/year (yr) for CDF and less than 1x10-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

b.

Other Changes that May Be Made Without Prior NRC Approval

1.

Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is

Document Control Desk - Attachment 11 LAR-16-01490 RC-18-0091 Page 3 of 4 functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:

"Fire Alarm and Detection Systems" (Section 3.8);

"Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9);

"Gaseous Fire Suppression Systems" (Section 3.10); and "Passive Fire Protection Features" (Section 3.11).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

2.

Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation dated February 11, 2015, to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

c.

Transition License Conditions

1.

Before achieving full compliance with 10 CFR 50.48(c), as specified by 2. and 3. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in 2. above.

Document Control Desk - Attachment 11 LAR-16-01490 RC-18-0091 Page 4 of 4

2.

The licensee shall implement the modifications to its facility, as described in Attachment S, Table S-1, "Plant Modifications Committed," of SCE&G letter RC-14-0196, dated December 11, 2014, as modified by Enclosure 1, Attachments 2 and 3, of SCE&G letter RC-18-0091, dated August 29, 2018. The licensee shall maintain appropriate compensatory measures in place until issuance of the safety evaluation.

3.

The licensee shall implement items listed in Attachment S, Table S-2, "Implementation Items," of SCE&G letter RC-14-0196, dated December 11, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by March 31, 2016 as follows:

a.

Items 3, 6, 7, 8, 10, 13, 14, 17, 19, and 21 within 180 days of NRC approval.

b.

Items 1, 2, 4, 11, and 12 by December 31, 2015.

c.

Items 5, 15, 16, 18, 20, 22, and 23 by March 31, 2016.

Document Control Desk LAR-16-01490 RC-18-0091 Page 1 of 12 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ENCLOSURE 2 CLARIFICATION AND EDITORIAL CORRECTIONS TO CURRENT NFPA 805 LICENSING BASIS

Document Control Desk LAR-16-01490 RC-18-0091 Page 2 of 12 1.0

SUMMARY

DESCRIPTION This enclosure addresses 3 topics not related to the de-scoped modifications:

Fire Doors IB105A and IB105B, the subject of Non-Cited Violation (NCV)05000395/2016001-01 (Reference 1) and Violation 05000395/2016007-01 (Reference 2).

Reactor Coolant Pump Lube Oil Collection System, the subject of Violation 05000395/2017002-01 (Reference 3), NCV 05000395/2017002-02 (Reference 3), and NCV 5000395/2013003-03 (Reference 4).

Clarification and Editorial Corrections to NFPA 805 Licensing Bases for VCSNS Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c).

2.0 FIRE DOORS IB105A AND IB105B DRIB/105A and DRIB/105B are Class A fire doors installed in a back-to-back configuration on the same frame within the three-hour fire rated barrier between the Heating, Ventilation, and Air Conditioning (HVAC) Water Chiller Pump A Room (Fire Zone IB07.01) and the HVAC A Water Chiller Room (Fire Zone IB23.01). Due to the unique construction of these doors, closing devices were not installed. The NEI 04-02 Table B-1 in the original NFPA 805 license amendment LAR-06-00055 (References 5 and 6) request did not refer to this fact.

This omission has been the subject of NRC inspection findings severity level IV NCV 05000395/2016001-01 (Reference 1) and Notice of Violation:

VIO 05000395/2016007-01 (Reference 2).

To address these violations, VCSNS has modified the doors to maintain them in the closed position. The doors cannot be physically opened without the aid of tools. VCSNS has analyzed the condition of fire doors IB105A and IB105B for adequacy. An engineering evaluation has been developed which evaluates the installation of fire doors IB105A and IB105B as acceptable. The evaluation is documented in TR0780E-006 Fire Protection Features Fire Protection Engineering Equivalency Evaluations. The NFPA 805 Chapter 3 code compliance table, currently maintained in the Fire Protection Design Basis Document (FP DBD) has been revised to reflect the above change:

Document Control Desk LAR-16-01490 RC-18-0091 Page 3 of 12 NFPA 805 Ch. 3 Reference Requirements / Guidance Compliance Statement Compliance Basis 3.11.3(1) [NFPA 80 - Fire Door and Window Requirements]

NFPA 80, Standard for Fire Doors and Fire Windows CE NRC The fire door design has been evaluated against the requirements of NFPA 80 Fire Doors and Windows, (1973).

Prior NRC approval of specialty doors (e.g. pressure and bullet resistant) were previously found to be acceptable and has been summarized in the defined evaluation.

Other evaluations also exist for miscellaneous door assemblies.

References Document ID DC0780D-007 - NFPA 805 Code Compliance-Passive Fire Protection Features TR0780E-006 - Fire Protection Features Fire Protection Engineering Equivalency Evaluations The NFPA 80 code compliance calculation DC0780D-007 has been revised and addresses the issue identified above.

The Reference Document ID for the applicable technical report identified in the original LAR was incorrectly identified as TR0787E-006. The correct report number is TR0780E-006.

3.0 REACTOR COOLANT PUMP (RCP) LUBE OIL COLLECTION SYSTEM Approval Request L12 was submitted in accordance with 10 CFR 50.48(c)(2)(vii) in RC-13-0142 (Reference 7) for NFPA 805 section 3.3.12(a) Reactor Coolant Pumps.

The approval request was for the potential of oil misting from the reactor coolant pumps due to normal motor consumption not captured by the oil collection system designed for pressurized and nonpressurized leakage and spillage.

Deficiencies associated with this approval request have been identified by NRC inspectors. The deficiencies include stating that the reactor coolant pump oil collection system is designed and was reviewed in accordance with 10 CFR 50, Appendix R, Section III.O to collect leakage from credible pressurized and nonpressurized leakage sites in the reactor coolant pump oil system, when in fact an NRC-identified Green NCV 05000395/2013003-03 (Reference 4), Failure to Adequately Design, Install and Maintain Oil Collection Devices for Reactor Coolant Pump Motors, had not been resolved. Additionally, the approval request used the word credible in place of potential when discussing leakage sites.

These deficiencies were captured as Violation 05000395/2017002-01 (Reference 3),

Failure to Implement Corrective Actions to Restore Compliance for Previous NRC-identified Green NCV 05000395/2013003-03 and Severity Level IV NCV 05000395/2017002-02 (Reference 3), Failure to Provide NRC Staff Complete and Accurate Information.

Document Control Desk LAR-16-01490 RC-18-0091 Page 4 of 12 The VCSNS reply to the NOV in a letter RC-17-0126 dated September 6, 2017 (Reference 8), included the following corrective step that will be taken:

ECR 50909, RCP Oil Enclosure, will be implemented during RF24. This ECR will ensure compliance with NFPA 805 for RCP oil collection requirements by modifying the oil lift pump enclosures, and fill and drain enclosures, and drip pans as necessary. This ECR will also address oil drips from the RCP exhaust duct to motor flange joint.

4.0 CLARIFICATION AND EDITORIAL CORRECTIONS TO NFPA 805 LICENSING BASES FOR VCSNS TRANSITION TO A RISK-INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(C)

Following NRC issuance of Amendment No. 199 to Renewed License No. NPF-12 (Reference 5 and 6) and its enclosed Safety Evaluation, an independent party was engaged to perform an assessment of VCS compliance with the NFPA 805 Licensing Bases. The results of this assessment included the identification of statements in the SE that are inconsistent with actual plant conditions. This was due to erroneous statements in the original LAR submittal or in response to RAIs. These items are described below and do not impact the plant or design bases.

4.1.1 Section 3.1.4.4, Bulk Gas Storage:

The assessment identified that the second paragraph of Safety Evaluation (SE)

Section 3.1.4.4, Bulk Gas Storage, should be clarified regarding the storage pressure of six hydrogen cylinders located outside the protected area south of the turbine building.

The text of Section 3.1.4.4 states that, the cylinders are maintained at 2400 psig. However, 2400 psig is the design pressure of the cylinders and is used to calculate their maximum total storage capacity. Actual plant conditions maintain the cylinders at a pressure less than 2400 psig, in accordance with standard operating procedures.

To accurately reflect plant conditions and operating procedures, the following clarification is proposed to the text in SE Section 3.1.4.4, second paragraph.

(Revision of wording from Reference 5 is shown in red text.)

In FPE RAI 10 (Reference 20), the NRC staff requested more detailed information regarding the extent of the hazard, the quantity and capacity of the tanks, a description of the refilling process and a description of the features that reduce the hazards associated with the alternative method.

In its response to FPE RAI 10 (Reference 8), the licensee stated that the bulk hydrogen storage facility consists of six hydrogen storage cylinders located outside the protected area south of the turbine building and that the cylinders are maintained at less than 2400 psig, and the tanks are filled at the storage location by a vendor using a tube trailer in accordance with site procedures. The licensee further stated that the tanks are surrounded by fencing and cement vehicle barrier posts and that there

Document Control Desk LAR-16-01490 RC-18-0091 Page 5 of 12 are no missile shields surrounding the tanks. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided adequate justification for the alternative method.

The proposed clarification above is to ensure correct understanding of the plant conditions and does not represent a change to the physical configuration or control of the plant.

4.1.2 Fire PRA Model The assessment identified that Safety Evaluation (SE) (Reference 5) Section 3.4.2.2, Fire PRA Model, paragraph 6 required a clarification on the number of scenarios provided in an additional response to PRA RAI 68 in reference 11.

In reference 11, it was stated that the total of 170 new scenarios were added to the Fire PRA model for the ungrouped transient zones. The 170 scenarios were listed in table 3 in reference 10. The assessment found that only 154 scenarios were listed in design calculation DC0780B-001, Fire Modeling: Generic Methodology.

Subsequent to the RAI response, 16 scenarios were determined not to be required, as a fire would not spread to adjacent transient zone via trays. Over time scenarios have been added and deleted during the Fire PRA maintenance process. The list of ungrouped transient zones scenarios is currently maintained in design calculation DC0780B-001, Fire Modeling: Generic Methodology.

4.1.3 Section 3.4.2.3.2, RAls Pertaining to Fire Modeling in Support of the VCSNS FREs:

A. The assessment identified that Safety Evaluation (SE) (Reference 5)

Section 3.4.2.3.2, RAls Pertaining to Fire Modeling in Support of the VCSNS Fire Risk Evaluations (FREs), required correction regarding the gap size assumptions used in fire modeling. The Safety Evaluation text states that the Hot Gas Layer (HGL) calculations assume an opening of 0.5 inch (one-half of an inch) below all doors, which is consistent with NUREG/CR-6850, for scenarios with closed doors and without mechanical ventilation. However, certain HGL calculations are not consistent with this statement.

The response to FM RAI 01.i incorrectly stated that for performance based hot gas layer calculations (DC0780F-series), the 0.5 inch opening corresponds to a leakage area fraction of 0.003, rather than the correct value of 0.006. A leakage area fraction of 0.003 corresponds to a 0.25 inch opening. The performance based fire modeling compliance strategy has been transitioned to a fire risk evaluation compliance strategy as part of this LAR submittal. The performance based fire modeling insights will no longer be used at VCSNS.

The response to FM RAI 01.i regarding the use of a 0.03 leakage area fraction for individual zone reports for HGL calculations (DC0780B-series) is accurate.

Document Control Desk LAR-16-01490 RC-18-0091 Page 6 of 12 The response to FM RAI 01.i also states several sensitivity cases were run for the hot gas layer CFAST calculations in which the equivalence ratio for the natural ventilation case was outside of the validation range (See DC0780B-366). In cases when the leakage area fraction was increased to bring the equivalence ratio into the validation range, the conclusions were unchanged.

It is requested that the SE be updated with this new information.

B. The assessment also identified that Safety Evaluation (SE) (Reference 5)

Section 3.4.2.3.2, RAls Pertaining to Fire Modeling in Support of the VCSNS Fire Risk Evaluations (FREs), required correction regarding calculations of sprinkler activation located near the ceiling. The Safety Evaluation texts states in response to FM RAI 01.t.iii, since sprinklers are not credited for activating until after 20 minutes, the licensee assumed that all targets within the initiating transient zone would be damaged and that sufficient smoke would be generated to allow for smoke detection and sprinkler activation for sprinklers outside the initial transient zone.

The Safety Evaluation text addressing FM RAI 01.t in 3.4.2.3.2 only addressed the information contained in the Oct 2012 RAI response RC-12-0142 (Reference 12) and did not include updated information that was submitted in a later response to FM RAI 01.01 from Oct 2013 (RC 0142). The later RAI response states:

The fire dynamic tools (FDT) for heat detector response time (Chapter 12, NUREG-1805) and sprinkler response time (Chapter 10, NUREG-1805) are no longer used to justify sprinkler activation in any of the fire zones.

For IB25.01.02, the fire scenarios that previously relied on the sprinkler response time FDTs (Chapter 10, NUREG-1805) have been revised. For three fixed ignition sources in fire zone IB25.01.02 (electrical cabinet DPN1HX and battery chargers XBC1X and XBC1X-2X), justification for sprinkler activation can be addressed using the approach already described in the previous RAI response to FM RAI 01.n, in which sprinkler heads located close to the ceiling are not credited until the ZOI targets are damaged. That is, the sprinklers will not be credited in protecting the tray nearest to the ignition source, but will be credited in preventing propagation after damage to the nearest trays.

The letter RC-12-0142 RAI response to FM RAI 01.n (Reference 12),

mentioned above, stated the following:

It should be noted that conservatism is achieved by assuming the ZOI targets are damaged before suppression starts. That is, there is no credit for suppression for the ignition source and targets within the ZOI. For example, in the approach 2 described earlier, the sprinkler will not be credited in protecting the nearby tray.

Document Control Desk LAR-16-01490 RC-18-0091 Page 7 of 12 However, it will be credited in preventing propagation after damage to the nearby tray is assumed.

It is requested that the SE be updated with this new information.

C. The assessment also identified in SE (Reference 5) Section 3.4.2.3.2, RAls Pertaining to Fire Modeling in Support of the VCSNS FREs, an issue that requires clarification, regarding the impact on the probability of abandonment in the MCR when considering transient fire heat release rates and average growth rates. The original response in letter RC-12-0142 (Reference 12) to FM RAI 01.g erroneously stated that:

The peak heat release rate for transients is 317 kW. Therefore, it was assumed that the CFAST simulations available would bound the results for the transient regardless of the faster growth rate as a peak of 317 kW is not expected to produce abandonment conditions based on CFAST results for the different ventilation conditions evaluated.

In actuality, it did not address the case of non-functioning HVAC for electrical cabinet fire average growth rates. For the case of a non-functioning HVAC system, the MCR abandonment time calculations show a heat release rate of approximately 197 kW is necessary to cause abandonment (DC0780B-100, Table 13), which is less than the 317 kW 98th percentile peak heat release rate for transient fires. For the 197 kW electrical cabinet fire, the CFAST results show time to abandonment of 12 minutes. For a transient fire, with a peak of 317 kW (98th percentile fire) and a time to peak of 8 minutes that is recommended by NUREG/CR-6850 Supplement 1 (page 17-2) for transient fires in a trash can, the time to reach 197 kW is calculated to be 6.3 minutes, using the following equation and solving for t:

(197 kW) / (317 kW) = (t/8 min)2 The current MCR abandonment calculations are based on electrical cabinet fires that have 12-minute growth rates. The MCR probability of abandonment calculations of non-suppression probability subtract 5 minutes from the calculated abandonment time. The 5 minutes is identified as time to suppression (DC0780B-100, Section 6.2); however, the latest guidance from NUREG/CR-6850 Supplement 1 recommends that the probability of non-suppression be calculated as time to damage (or in this case, time to abandonment) minus time to detection (which is zero for the continually manned control room). Therefore, the current calculation of time to abandonment, based on a time to peak HRR of 12 minutes and reduced by a 5 minute time to suppression, is similar to the time to abandonment calculation for the case of a time to peak HRR of 7 minutes, without the 5-minute time to abandonment reduction. A comparison of the non-suppression probability for the electrical fire with suppression after 7 minutes and a transient fire with suppression after 6.3 minutes shows a value of 0.10 for the electrical cabinet and 0.13 for the transient fire.

Document Control Desk LAR-16-01490 RC-18-0091 Page 8 of 12 A sensitivity calculation in the UNC notebook (DC00340-009) shows that calculating the probability of abandonment using the transient fire heat release rate, rather than the electrical cabinet heat release rate has a small impact on the results: an increase of 1E-8 in CDF and a decrease of 2E-11 in LERF.

It is requested that the SE be updated with this new information.

4.1.4 Section 3.5.1.8, Plant Fire Barriers and Separations:

The assessment identified that Safety Evaluation (SE) (Reference 5) Section 3.5.1.8, Plant Fire Barriers and Separations, required an update regarding underground electrical duct banks near the Diesel Generator building. This distance was erroneously provided in RC-14-0027 (Reference 9) Attachment C under FPEEE-DB-01. The SE makes the statement that single channel duct banks within Fire Area DB are isolated from one another by 30 feet. Contrary to this statement, actual spacing of the duct banks varies.

It was identified that spacing between 1A and 1B single channel duct banks varies in Fire Area DB. In some instances, the edge-to-edge separation is less than 20 feet, but greater than 10 feet. The varied separation was evaluated and determined to have no impact on the Fire Risk Evaluation, as the duct banks are encased in concrete and buried. The soil and concrete provide an adequate level of protection against fire damage to both channels.

To address fire modeling calculation input, the following revisions to the Safety Evaluation are proposed for Section 3.5.1.8, third paragraph. (Revision of wording from Reference 5 is shown in red text.)

LAR Table 4-3, "NSCA FPEEs/Licensing Actions," as supplemented, documents the FPEEEs, the licensing action and a brief description. LAR Attachment K "Existing Licensing Action Transition" lists the Fire Area, Licensing Action Number, whether it is to transition to NFPA 805 and pertinent NFPA 805 comments. Additionally, the type of deviation (i.e.,

lack of automatic suppression, lack of 1-hour fire rated barrier), detailed technical information, and a basis are provided for each licensing action.

Similar information to LAR Attachment K can also be found in LAR Attachment C, "NEI 04-02 Table B-3 Fire Area Transition" on a fire area basis. These documents identify passive fire protection features that are required to meet NFPA 805 criteria. Fire Area DB, underground duct bank, credits conduits within duct banks that are encased in at least 5 inches of concrete, 6 inches of concrete that separate the channels of redundant circuits, and single channel duct banks are isolated from one another by 30' at least 10 feet of concrete and soil. Embedded conduits are also credited in fire area DB. Fire area IB07, IB Chilled Water Pump Rooms 412, credits 3-hour fire barriers, between the chilled water pump rooms from adjacent areas and existing 1-hour rated radiant energy shields. Fire area MH02, B Train of MH02, in the manhole, credits 6 inches of concrete wall (containing drainage opening at bottom of wall) between redundant circuits with a 2-foot thick concrete cover. SE Section

Document Control Desk LAR-16-01490 RC-18-0091 Page 9 of 12 3.1 provides the results of the NRC staff's evaluation of the acceptability of fire barriers and separations against the NFPA 805 Chapter 3 Section 3.11 minimum design requirements for these fire protection features.

4.1.5 Table 3.5-2, Previously Approved Licensing Actions Being Transitioned:

The assessment identified a discrepancy in Safety Evaluation (SE) Table 3.5-2, based on inaccurate information provided in LAR-06-00055 (Reference 6)

Attachment K, Existing Licensing Action Transition. Specifically, the discrepancy in the NFPA 805 licensing basis was initiated in the details and basis of LA-IB25-02 of LAR-06-00055 Attachment K, as described below.

SCE&G requested a deviation from Appendix R III.G.2 criteria for the Intermediate Building (IB) Fire Area IB25 in a [[letter::05000395/LER-1985-013, :on 850429,reactor Trip Occurred Due to lo-lo Steam Generator Water Level in Steam Generator B. Caused by Transients in Deaerator Tank Level & Main Feedwater Pump Discharge Pressure|letter dated May 29, 1985]] (ADAMS Legacy No. 8506050222). This deviation was later withdrawn, and a new deviation was requested for the same Fire Area in a SCE&G letter dated April 23, 1986 (ADAMS Legacy No. 8604280179). The NRC approved this deviation request in a Safety Evaluation (SE) (Reference 10) dated July 27, 1987.

LAR-06-00055, Attachment K was based on the original deviation request that was withdrawn and not the second deviation request for Fire Area IB25.

(Updated Attachment K information for Licensing Action LA-IB25-02 is included below.) Proposed corrections to LA-IB25-02 in the SE Table 3.5.2 are provided below, for Reference. (Revisions shown in red text.)

Additionally, the erroneous reference to the original deviation request contributed to several errors in the description of engineering evaluation FPEEE-IB25-03 included in the LAR-06-00055 Attachment C. Corrections have been incorporated into the updated Attachment C provided in Enclosure 1 of this submittal.

Document Control Desk LAR-16-01490 RC-18-0091 Page 10 of 12 The following corrections are provided to LA-IB25-02 information, previously provided in LAR-06-00055, Attachment K. (Revisions shown in red text.)

LA-IB25-02 Transition to 805? Yes 805 Comments: This Licensing Action is credited in the NSCA and is to be transitioned into NFPA 805.

Appendix R Deviation, Intermediate Building - Lack of 1-hour fire rated barrier (III.G.2.c criteria)

Details:

Redundant trains of cables for the Service Water (SW) Booster Pump are separated by an unrated floor/ceiling assembly containing unsealed penetrations and unrated access hatches. Train A equipment and cables are located on IB elevations 412 (Fire Zones IB25.01.01 and IB25.01.02) and 436 (Fire Zone IB25.06). Train B equipment and cables are located on IB elevation 412 (Fire Zone IB25.01.02). The A Train cables on El. 412 are provided with a 1-hour ERFBS. The A Train cables have not been wrapped on El. 436. Automatic fire suppression and detection is provided on the 412 elevation. Detection is provided at elevation 436. (Note that automatic suppression is provided in a small portion of the 436 elevation but is not provided throughout other fire zones in the Fire Area, including the area of concern for LA-IB25-02.)

Redundant trains of SW Booster Pump required support circuits are separated horizontally by 12-ft and by a reinforced concrete wall with unprotected openings. IB-25.1 - Train A equipment and cables. IB-25.10 - Train B power and control cables for the DG (causes loss of onsite power to Train B SW Booster Pump).

Basis:

A Deviation request per the 5/29/1985 4/23/1986 SCE&G submittal provides the following justification for the lack of a 1-hour fire rated barrier throughout the Fire Area as required by Section III.G.2.c of Appendix R. This deviation was accepted by the NRC in a letter dated 7/27/1987:

• Redundant circuits are separated vertically horizontally by 12-ft and by a reinforced concrete floor/ceiling assembly wall with unprotected openings.

• Automatic suppression and detection in Fire Zone IB25.01 at El. 412

• Automatic detection in Fire Zone IB25.01 at El. 412 and IB25.06 at El. 436.

• 1-hr ERFBS provided for Train A cables at El. 412.

• Penetrations in the floor/ceiling assembly are sealed for the SW Booster Pump Train A cables, and for penetrations directly above the redundant Train B cables.

• Manual fire suppression is provided by interior manual hose stations and portable extinguishers.

• Automatic detection in Train B cable chase

• 3-hr fire barrier with unprotected openings around Train B cable chase FPEEE

Reference:

Post-transition bases for acceptability, see TR0780E-001, Attachment IB25-03

Document Control Desk LAR-16-01490 RC-18-0091 Page 11 of 12 To reflect the correct information for LA-IB25-02, the following revisions to the Safety Evaluation are proposed for Table 3.5.2. (Revisions shown in red text.)

Licensing Action Description Applicable Fire Areas

NRC Staff Evaluation

LA-IB25 Appendix R Deviation, Intermediate Building - Lack of 1-hour fire rated barrier (III.G.2.c criteria)

IB25 The basis for approval as described by the licensee in LAR Attachment K is redundant circuits are separated vertically by a reinforced concrete floor/ceiling assembly with unprotected openings.

Automatic suppression is provided in Fire Zone IB25.01 at El. 412. Automatic detection is provided in Fire Zone IB25.01 at El. 412 and IB25.06 at El. 436. 1-hr ERFBS is provided for Train A cables at El. 412. Penetrations in the floor/ceiling assembly are sealed for the SW Booster Pump Train A cables, and for penetrations directly above the redundant Train B cables.

Manual fire suppression is provided by interior manual hose stations and portable extinguishers.

horizontally by 12-ft and a reinforced concrete wall with unprotected openings. Automatic suppression and detection is installed in fire zone IB25.01 and automatic detection is installed in the Train B cable chase. Also there is a 3-hr fire barrier with unprotected openings around Train B cable chase.

Based on the previous staff approval of this deviation in an SE dated 7/27/87 (Reference 32),

and inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.

Document Control Desk LAR-16-01490 RC-18-0091 Page 12 of 12

5.0 REFERENCES

1.

NRC letter dated May 11, 2016, Virgil C. Summer Nuclear Station - NRC Problem Identification and Resolution Inspection Report 05000395/2016001 and Notice of Violation (ML16132A228)

2.

NRC letter dated November 22, 2016, Virgil C. Summer Nuclear Station - NRC Problem Identification and Resolution Inspection Report 05000395/2016007 and Notice of Violation (ML16327A378)

3.

NRC, Letter dated August 9, 2017, Virgil C. Summer Nuclear Station, Unit 1 -

NRC Integrated Inspection Report 05000395/2017002 and Notice of Violation (ML17221A115)

4.

NRC, Letter dated August 6, 2013, Virgil C. Summer Nuclear Station, Unit 1 -

NRC Integrated Inspection Report 05000395/2013003 and Notice of Violation (ML13218A334)

5.

Shawn Williams, NRC, Letter to Thomas Gatlin, Virgil C. Summer Nuclear Station, Unit 1 - Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (TAC No. ME7586), dated February 11, 2015. (ML14287A289)

6.

SCE&G letter dated November 15, 2011, License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition), (RC-11-0149) (ML14063A455)

7.

SCE&G letter dated October 14, 2013, License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, (RC-13-0142) (ML13289A194)

8.

SCE&G letter dated September 6, 2017, Virgil C. Summer Nuclear Station (VCSNS), Unit 1 Docket No. 50-395, Operating License No. NPF-12, Reply to Notice of Violation, (RC-17-0126) (ML17249A663)

9.

SCE&G letter dated February 25, 2014, License Amendment Request-LAR 00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, (RC-14-0027) (ML14063A455)

10.

Jon Hopkins, NRC, Letter to D. A. Nauman, Virgil C. Summer Nuclear Station -

Appendix R Reanalysis (TAC No. 57853), dated July 27, 1987 (ML15030A099)

11.

Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No.

50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated April 1, 2013 (ML13092A333).

12.

Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No.

50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated October 10, 2012 (ADAMS Accession No. ML12297A218).

Document Control Desk LAR-16-01490 RC-18-0091 Page 1 of 1 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ENCLOSURE 3 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by SCE&G, Virgil C. Summer Nuclear Station in this document. Any other statements in this submittal are provided for information purposes and are not considered to be commitments. Please direct questions regarding these commitments to Bruce L. Thompson, Manager, Nuclear Licensing, (803) 931-5042.

Commitment Due Date/Event NO NEW COMMITMENTS WERE GENERATED.