BSEP 14-0092, Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805 (NRC TAC ME9623 and ME9624)

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Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805 (NRC TAC ME9623 and ME9624)
ML14234A326
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 08/15/2014
From: Hamrick G
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 14-0092, TAC ME9623, TAC ME9624
Download: ML14234A326 (14)
v • d • e


Text

George T. Hamrick Letter Enclosures 3, 4, and 6 Contain Vice President ENER~WGY. Security-Related Information - Brunswick Nuclear Plant P.O. Box 10429 Withhold in Accordance with 10 CFR 2.390 Southport, NC 28461 o: 910.457.3698 August 15, 2014 Serial: BSEP 14-0092 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805 (NRC TAC Nos. ME9623 and ME9624)

References:

1. Letter from Michael J. Annacone (Carolina Power & Light Company) to U.S.

Nuclear Regulatory Commission (Serial: BSEP 12-0106), License Amendment Request to Adopt NFPA 805 Performance-BasedStandard for Fire Protection for Light Water Reactor Electric GeneratingPlants (2001 Edition), dated September 25, 2012, ADAMS Accession Number ML12285A428

2. Letter from Michael J. Annacone (Carolina Power & Light Company) to U.S.

Nuclear Regulatory Commission (Serial: BSEP 12-0140), Additional Information Supporting License Amendment Request to Adopt NFPA 805 Performance-BasedStandardfor Fire Protection for Light Water Reactor Electric GeneratingPlants (2001 Edition), dated December 17, 2012, ADAMS Accession Number ML12362A284

3. Letter from Farideh Saba (USNRC) to George T. Hamrick (Duke Energy Progress, Inc.), Second Request for Additional Information Regarding Voluntary Risk Initiative National Fire ProtectionAssociation Standard805 (TAC Nos. ME9623 and ME9624), dated February 12, 2014, ADAMS Accession Number ML14028A178
4. Letter from George T. Hamrick (Duke Energy Progress, Inc.) to U.S. Nuclear Regulatory Commission (Serial: BSEP 14-0029), Response to Additional Information Regarding Voluntary Risk Initiative NationalFire Protection Association Standard805 (NRC TAC Nos. ME9623 and ME9624), dated March 14, 2014, ADAMS Accession Number ML14079A233
5. Letter from George T. Hamrick (Duke Energy Progress, Inc.) to U.S. Nuclear Regulatory Commission (Serial: BSEP 14-0035), Response to Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard805 (NRC TAC Nos. ME9623 and ME9624), dated April 10, 2014, ADAMS Accession Number ML14118A105 When Enclosures 3, 4 and 6 are removed, this document is no longer Security-Related

U.S. Nuclear Regulatory Commission Page 2 of 4

6. Letter from George T. Hamrick (Duke Energy Progress, Inc.) to U.S. Nuclear Regulatory Commission (Serial: BSEP 14-0076), Response to Additional Information Regarding Voluntary Risk Initiative NationalFire Protection Association Standard805 (NRC TAC Nos. ME9623 and ME9624), dated June 26, 2014, ADAMS Accession Number ML14191A672
7. Electronic Mail from Andrew Hon (U.S. Nuclear Regulatory Commission) to William R. Murray (Duke Energy Progress, Inc.), Brunswick Steam Electric Plant, Units I and 2 - Request for Additional Information Regarding Voluntary Risk Initiative NationalFire Protection Association Standard 805 (TAC Nos. ME9623 and ME9624), dated July 24, 2014, ADAMS Accession Number ML14205A592 Ladies and Gentlemen:

By letter dated September 25, 2012 (i.e., Reference 1), as supplemented by letter dated December 17, 2012 (i.e., Reference 2), Duke Energy Progress, Inc., submitted a license amendment request (LAR) to adopt a new, risk-informed, performance-based (RI-PB) fire protection licensing basis for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

On February 12, 2014 (i.e., Reference 3), the NRC provided a request for additional information (RAI) regarding the fire probabilistic risk assessment. By letters dated March 14, 2014 (i.e.,

Reference 4); April 10, 2014 (i.e., Reference 5); and June 26, 2014 (i.e., Reference 6), Duke Energy responded to the RAI. Subsequently, on July 24, 2014 (i.e., Reference 7), the NRC provided an electronic, follow-up RAI. The response to the follow-up RAI is provided in of this letter.

As communicated in a telephone call with NRC representatives on August 6, 2014, Duke Energy has determined that several modifications originally planned for completion as part of the transition to 10 CFR 50.48(c), NFPA 805 Fire Protection licensing basis are no longer needed. Further information regarding the elimination of these modifications is provided in .

This document contains no new regulatory commitments.

Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

U.S. Nuclear Regulatory Commission Page 3 of 4 I declare, under penalty of perjury, that the foregoing is true and correct. Executed on August 15, 2014.

Sincerely, George T. Hamrick

Enclosures:

1. Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805
2. Discussion on Elimination of Selected Modifications and Implementation Items
3. Updated Final License Amendment Request Attachment S, Modifications and Implementation Items (Security-Related Information - Withhold from Public Disclosure)
4. Basis for Modifications Removed from License Amendment Request Attachment S, Modifications and Implementation Items (Security-Related Information - Withhold from Public Disclosure)
5. Replacement Pages for License Amendment Request Attachment B, NEI 04-02 Table B Nuclear Safety CapabilityAssessment - Methodology Review
6. Replacement Pages for License Amendment Request Attachment C, NEI 04-02 Table B Fire Area Transition (Security-Related Information - Withhold from Public Disclosure)

U.S. Nuclear Regulatory Commission Page 4 of 4 cc (with all Enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Victor M. McCree, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U. S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission ATTN: Ms. Michelle P. Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 cc (with Enclosure 1 only):

Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only)

Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1645 lee.cox@dhhs.nc.gov

Enclosure 1 Page 1 of 6 Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805 By letter dated September 25, 2012, as supplemented by letter dated December 17, 2012, Duke Energy Progress, Inc., submitted a license amendment request (LAR) to adopt a new, risk-informed, performance-based (RI-PB) fire protection licensing basis for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

On February 12, 2014, the NRC provided a request for additional information (RAI) regarding the fire probabilistic risk assessment. By letters dated March 14, 2014; April 10, 2014; and June 26, 2014, Duke Energy responded to the RAI. Subsequently, on July 24, 2014, the NRC provided an electronic, follow-up RAI.

Duke Energy's response to the RAI is provided below.

Probabilistic Risk Assessment (PRA) Request for Additional Information (RAI) 01.f.ii.03 In a letter dated June 26, 2014 (Agencywide Document Access and Management System (ADAMS) Accession No. ML14191A672) the licensee responded to PRA RAI 01.f.ii.02 and stated that to "assess the risk due to loss of control, the applicable bin-specific ignition frequency was apportioned to each ignition source in the main control room." The term "loss of control" implies to NRC staff that failure of main control room (MCR) functionality from fire-induced failures, including ones due to fires outside the MCR, leads to the need to abandon the MCR and perform alternate shutdown. Given previous statements that Brunswick Steam Electric Plant (BSEP) does not credit MCR abandonment upon loss of control, please explain what the term "loss of control" means in the cited statement. Please confirm that loss of functionality of components, instruments, etc., in the MCR due to fires outside the MCR are included in the risk and delta-risk estimates.

Response

As used in the cited statement in the response to PRA RAI 01 .f.ii.02, "loss of control" always represents the same general condition - that the plant cannot be brought to a safe state from the MCR.

PRA RAI 01 .f.ii.02 concerned the treatment of the MCR abandonment for MCR fires. Use of the term "loss of control" within this context was not intended to imply that loss of control was not quantified for fires outside thp MCR. Fires both inside and outside the MCR contribute to the risk and delta-risk estimates for loss of control. The loss of functionality of components, instruments, etc., in the MCR due to fires outside the MCR is confirmed to be included in the risk and delta-risk estimates.

PRA RAI 01.f.ii.04 In a letter dated June 26,2014 (ADAMS Accession No. ML14191A672) the licensee responded to PRA RAI 01 .f.ii.02 and stated that their "ASSD [Alternate Safe Shutdown] strategy is not affected by fire-induced equipment failures in the main control room." Please confirm that BSEP's ASSD strategy and modeling of that strategy for the Fire PRA accounts for fire-induced equipment impacts in the MCR, including potential spurious operations, particularly ones that may not be "recoverable," if needed, upon transfer of control to the ASSD panel. If modeling of BSEP's ASSD strategy does not account for fire-induced MCR equipment impacts on success

Enclosure 1 Page 2 of 6 of ASSD, including spurious operations as cited above, then address these impacts as part of the integrated analysis performed in response to PRA RAI 23 and, to the extent applicable PRA RAI 24.

Response

In the response to PRA RAI 01 .f.ii.02, the ASSD strategy was discussed within the limited context of how operator recovery was credited during Main Control Room Abandonment (MCRA) for habitability. In assessing the risk of MCRA for habitability, the operator recovery is based, in part, on the Human Error Probability of failing to implement the mitigating action in the ASSD procedures. The assessment does not depend on whether fire-induced spurious operations exist which require that action. Instead, fire-induced spurious operations and core damage were assumed to exist if the required action was not taken. For example, core damage was assumed if the operator fails to remove DC power from the DC solenoids of the Main Steam Isolation Valve (MSIV) air lines without regard to whether the fire actually caused spurious opening of the MSIVs. This conservative treatment in a limited context results in little credit (i.e., 0.253) for recovery of MCRA due to habitability.

The response to PRA RAI 01 .f.ii.02 also described how the frequency for MCR ignition sources was conservatively "double counted" (e.g., once to quantify the associated risk as a "loss of control" scenario and again for a loss of habitability contribution). The "loss of control" scenarios accounted for the risk associated with fire-induced equipment impacts, including potential spurious operation whether in the MCR or otherwise, and were not credited with operator recovery of implementing the ASSD procedures.

For fire-induced equipment impacts beyond those postulated for the ASSD strategy, a systematic review of spurious and multiple spurious operation (MSO) scenarios was performed using an Expert Panel. The purpose for this MSO Expert Panel was to identify potential MSO scenarios that could place the plant in an unrecoverable condition, or result in unrecoverable damage to required equipment, and to determine which scenarios were credible and may need to be incorporated into the Safe Shutdown Analysis (SSA) and Fire PRA models. The MSO Expert Panel considered MSOs identified in industry guidance, by other expert panels, and from their own initiative.

The evaluation of particular MSO cases and disposition with regard to modeling in the Fire PRA was documented in Attachments 3 and 4 of BNP-PSA-085, BNP Fire PRA - Component Selection. This included MSOs (e.g., cases El1-5d, E21-ri-1, and P41-5c) postulated to occur during the time it takes to evacuate the MCR and prior to isolating the circuits and taking component control remotely. Those scenarios that result in core damage would be characterized as a loss of control, but the associated risk would not be mitigated by crediting any recovery for implementation of ASSD procedures.

PRA RAI 25 The recent License Event Report (LER) 14-004-00, "Fire Related Unanalyzed Condition that Could Impact Equipment Credited in Safe Shutdown Analysis," (ADAMS Accession No. ML14149A244) stated that:

On March 20, 2014, as a result of the transition process from 10 CFR 50, Appendix R, to NFPA 805, a review of the Brunswick Steam Electric Plant Safe Shutdown Analysis determined that a postulated fire in specific fire areas could disable critical components,

Enclosure 1 Page 3 of 6 potentially resulting in equipment required for safe shutdown being inoperable. The safety significance of this event is minimal. Deterministic analysis methods used to comply with Appendix R require every possible fire scenario to be addressed; however, the risk posed by these hypothetical events has been determined by analysis to be minimal ... This condition is being reported per 10 CFR 50.73(a)(2)(ii)(B) as an event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety ... The safety significance of this event is minimal.

Fire watches had previously been established in affected areas prior to the time of discovery for reasons other than this event. The conditions identified here are based on hypothetical fire scenarios that have not actually occurred. A probabilistic safety assessment developed to analyze this event shows that the core damage frequency (CDF) and large early release frequency (LERF) are less than red per the significance determination process ... Applicable procedures will be revised by July 31, 2014, to prescribe the required actions for mitigating the effects of a fire in the affected areas.

This represents the discovery of new fire-induced accident scenarios that may not have been modeled in the Fire PRA.

a. Please explain whether the conditions described in LER 14-004-00 could impact fire core damage frequency (CDF), large early release frequency (LERF), change in (A)

CDF, or ALERF.

b. Please explain whether these conditions are currently addressed in the Fire PRA.
c. If the conditions described by LER 14-004-00 are not currently addressed in the Fire PRA, please provide justification, or update the Fire PRA to include these scenarios.

If this update is performed prior to transition, please ensure that the responses to PRA RAI 23 and PRA RAI 24 capture the effects of the update. This request can also be addressed by an alternative sensitivity study that incorporates the anticipated effects of the model update and add an implementation item to update the model post-transition, including an assessment of the potential need for a focused-scope peer review if this update qualifies as a PRA "upgrade" as defined by the ASME/ANS PRA Standard. If this option is selected, please include the results of the sensitivity in the aggregate analysis described by PRA RAI 23 and, to the extent applicable PRA RAI 24.

Response

As described in LER 14-004-00, a detailed revalidation of the BSEP Safe Shutdown Analysis (SSDA) identified the following six conditions that may not ensure a protected train of equipment remains available under certain postulated fire scenarios:

1. To ensure net positive suction head (NPSH) requirements are met for Residual Heat Removal (RHR) system pumps, Containment Over Pressurization (COP) must be maintained. This is accomplished by ensuring Reactor Building Closed Cooling Water (RBCCW) pumps remain off. Analysis determined that postulated fires in areas CB-23E, RB1-N, and RB1-S, located in the MCR and in the Reactor Building, could have affected the ability to secure the RBCCW pumps and keep them secured. This potentially affects Unit 1.

Enclosure 1 Page 4 of 6

2. A postulated fire in areas TB1 and RB1-S, located in the Unit 1 Turbine Building and in the southern half of the Unit 1 Reactor Building, could potentially damage cables that control the supply fans in the Diesel Generator Building. This could disable the fans which are assumed operable in the SSDA. This potentially affects both units.
3. A postulated fire in area DG-07, located in the switchgear room for 480V bus E6, could potentially lead to the control power cables for the under voltage circuitry being spuriously energized. This could lead to a spurious signal being sent to the El switchgear controls that would interfere with the operation of cross-tie breakers. A fire in this area could also potentially damage the normal control power cables for bus E6 and disable the protective relaying function needed for transmitting power from bus E4 to bus E2. This potentially affects both units.
4. A postulated fire in area DG-16E, located in the supply air plenum and exhaust fan area of the 50 foot elevation of the Diesel Generator Building, could interrupt ventilation needed for the Division I switchgear to maintain long term control power to the Start-up Auxiliary Transformer. This potentially affects both units.
5. A postulated fire in area RB2-N, located in the northern half of the Unit 2 Reactor Building, could blow the fuse for the control circuitry for breakers that are fed from the El bus, thus interrupting their control power. This would prevent the bus from shedding its loads which would then prevent them from being sequenced back onto the bus after it is powered by the Emergency Diesel Generators. A postulated fire in this area could also blow the fuse for the control circuitry for breakers that are fed from bus E4, interrupting control power for loads fed from this bus. This would prevent the bus from shedding its loads which would then prevent them from being sequenced back onto the bus after it is powered by the Emergency Diesel Generators. This potentially affects both units.
6. To ensure COP to meet NPSH requirements for the Unit 2 RHR system pumps, RBCCW pumps must remain off. Analysis determined that postulated fires in areas CB-23E, RB2-N, and RB2-S, located in the MCR and in the Reactor Building, could have affected the ability to secure the RBCCW pumps and keep them secured. This potentially affects Unit 2.

However, these conditions do not necessarily represent new fire-induced accident scenarios that were not modeled in the Fire PRA (i.e., Reference EC 97680). As further described in the LER, these conditions were based on hypothetical fire scenarios for which the safety significance was determined to be minimal.

Condition #1:

a. CDF and LERF are not impacted because the Fire PRA does not credit COP to provide NPSH for RHR pumps; consequently ACDF and ALERF are likewise not impacted.
b. This is addressed in the thermal hydraulic analyses supporting the modeling logic in the Fire PRA.
c. Neither additional justification nor an update to the Fire PRA is applicable.

Enclosure 1 Page 5 of 6 Condition #2:

a. For the cables in the Turbine Building, CDF, LERF, ACDF, and ALERF are not impacted because the cables of concern are not a target for any fire in the Turbine Building. For the cables in the Reactor Building, CDF and LERF are not impacted because the cables of concern are already mapped to fail the associated EDG; likewise ACDF and ALERF are not impacted because the compliant case already included the effect of these cables as variances from deterministic requirements (VFDRs).
b. This is addressed in the development of fire scenarios for the Fire PRA.

C. Neither additional justification nor an update to the Fire PRA is applicable.

Condition #3:

a. CDF and LERF are not impacted because the cables of concern are already mapped to fail 1-E6, 1-El and control power to 1-E2 (consistent with the Internal Events PRA, the Fire PRA does not credit the cross-divisional or 3-bus crosstie of 2-E4 through 1-E2 to restore 1-El); likewise ACDF and ALERF are not impacted because the compliant case already included the effect of these cables as VFDRs.
b. This is addressed in the development of fire scenarios for the Fire PRA.

C. Neither additional justification nor an update to the Fire PRA is applicable.

Condition #4:

a. CDF, LERF, ACDF, and ALERF are not impacted because this condition concerns the long-term effects on equipment life due to a post-fire failure to restore ventilation which was interrupted by the detection of the fire, although such restoration is in the fire response procedures as a post-fire action.
b. This condition is not addressed in the Fire PRA.

C. This condition would not be addressed in the Fire PRA, because:

  • This condition does not concern fire-induced equipment failure within the mission time of the Fire PRA. Instead, the success criteria for this condition were based on ensuring indefinite operation with no reduction in equipment life.

0 An event tree analysis, based on the scenario event frequencies (i.e., with no suppression credit) and the projected Human Error Probability for this long-term action, indicated no reason to include this condition in a more fully developed fault tree for the Fire PRA beceuse the associated cutsets (i.e., assuming post-fire equipment damage) would be below truncation even without consideration for other mitigating equipment.

Condition #5:

a. CDF and LERF are not impacted because the cables of concern are already mapped to fail the associated EDG due to a failure to load shed; likewise ACDF and ALERF are not impacted because the compliant case already included the effect of these cables as VFDRs.
b. This is addressed in the development of fire scenarios for the Fire PRA.

C. Neither additional justification nor an update to the Fire PRA is applicable.

Enclosure 1 Page 6 of 6 Condition #6:

a. CDF and LERF are not impacted because the Fire PRA does not credit COP to provide NPSH for RHR pumps; ACDF and ALERF are likewise not impacted.
b. This is addressed in the thermal hydraulic analyses supporting the modeling logic in the Fire PRA.
c. Neither additional justification nor an update to the Fire PRA is applicable.

Based on the response above, no change to the response for PRA RAI 23 or for PRA RAI 24 has been proposed.

Enclosure 2 Page 1 of 2 Discussion on Elimination of Selected Modifications and Implementation Items As communicated in a telephone call with NRC representatives on August 6, 2014, Duke Energy has determined that several modifications originally planned for completion as part of the transition to 10 CFR 50.48(c), NFPA 805 Fire Protection Licensing Basis, are no longer needed.

The License Amendment Request (LAR) submittal included an Attachment S, "Modification and Implementation Items." This attachment identified the modifications (i.e., Engineering Changes (EC)) agreed to be completed as part of transition to the 10 CFR 50.48(c), NFPA 805 Fire Protection licensing basis. As part of the implementation and development of the transition, a number of these EC's were identified as no longer being needed to meet the assumptions used in the analysis. As such, Duke Energy is deleting the modifications listed below from the final LAR Attachment S.

  • Item 6: Provide modification to change the valves to a fail-safe condition of closed on loss of valve support for Hotwell Makeup Valves, 1-CO-LV-1-2, 2-CO-LV-1-2.
  • Item 7: Provide separation / protection of the following cables in the Main Control Room:

1JG8-WI3, 1AQ6-14B, 1AQ6-IJ1, 2AQ6-HZ3.

" Item 10: Evaluate and modify valves, as necessary, to address pressure boundary concerns due to fire induced spurious actuations. Perform a study for the extent of condition for valves of concern.

" Item 11: Provide enhanced risk margin for Core Damage Frequency (CDF)/Large Early Release Frequency (LERF) for the Main Control Room by modification of suspended ceiling configuration to allow for an effective increase in ceiling height and associated volume of the Main Control Room. of this letter provides a final updated copy of the LAR Attachment S that shows the requested modifications as "deleted." This Attachment S should be referred to when developing the final Safety Evaluation documentation. The final updated Attachment S (i.e., Enclosure 3) supersedes, in its entirety, the Attachment S previously provided as part of Duke Energy's letter dated March 14, 2014 (i.e., Reference 4 of the cover letter). provides a list of the modifications being eliminated, with a brief problem statement and conclusion statement that explains the basis for not completing each modification. In addition, the complete EC evaluation packages have been posted to the Brunswick NFPA 805 SharePoint location for detailed review.

Key to removal of the modifications from Attachment S is the fact that no conditions or assumptions previously submitted in support of the Fire Probabilistic Risk Assessment would be altered by removal of the modifications. This is further discussed in the EC evaluations supporting the elimination of each modification. As such, there are no changes in the previously reported values for CDF, LERF, ACDF, or ALERF given the absence or removal of the identified Modifications (EC's). Fire Safety Analyses (FSAs) for the individual fire area did not consider these modifications as Defense-in-Depth attributes when determining overall capability to meet performance criteria.

Enclosure 2 Page 2 of 2 A systematic review was conducted of the BSEP LAR and supporting materials, and revisions have been incorporated for those LAR sections and items where reference was made to deleted modifications. Enclosures 5 and 6 provide the replacement pages for LAR Attachment B, NEI 04-02 Table B Nuclear Safety CapabilityAssessment - Methodology Review, and LAR Attachment C, NEI 04-02 Table B Fire Area Transition, respectively, to reflect elimination of the modifications. Fire Area specific FSAs (i.e., RW-01 and CB-23E) will be updated to delete reference to the modifications (i.e., reference new Table S-2 in Enclosure 3, Implementation Item 16).

Enclosure 5 Replacement Pages for License Amendment Request Attachment B, NEI 04-02 Table B Nuclear Safety CapabilityAssessment - Methodology Review (1 Page)

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Reference Document Doc Detail BNP-E-9,004, Safe Shutdown Analysis Report Section 3.B.1 BNP-E-9,010, Safe Shutdown Analysis In Case of Fire Section 3.1 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 9.3.3 and 9.3.4 (NSCA)

Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supportina Detail B2-GAP-004 Closed 04-27-12 El VFDR Disposition EC 85096R0 is evaluating all applicable IMOVs for 92- The EVAL-EC has concluded that none of the motor 18 issues, inlcuding those thay may be pressure operated valves subject to spurious operation due to boundary concerns. fire induced cable damage constitute pressure boundary concerns; therefore, no modifications or further evaluations are warranted.

FRE/Change Eval/Mod C59 Corrective Action

Reference:

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected BSEP LAR Rev 1 Page B-78

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