05000324/LER-2014-004, Regarding Fire Related Unanalyzed Condition That Could Impact Equipment Credited in Safe Shutdown Analysis

Regarding Fire Related Unanalyzed Condition That Could Impact Equipment Credited in Safe Shutdown Analysis

ML14149A244
Person / Time
Site:	Brunswick
Issue date:	05/16/2014
From:	Hamrick G Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 14-0052 LER 14-004-00
Download: ML14149A244 (6)

LER-2014-004, Regarding Fire Related Unanalyzed Condition That Could Impact Equipment Credited in Safe Shutdown Analysis

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3242014004R00 - NRC Website

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DUKE George TA.

Vice P ENERGY9 Brunswick Nucle Duke Energy P P.O. Bo Southport, N(

o: 910.4 10 CFR 50.73 MAY 16 2014 Serial: BSEP 14-0052 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

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 Licensee Event Report 1-2014-004 In accordance with the Code of Federal Regulations, Title 10, Part 50.73, Duke Energy Progress, Inc., submits the enclosed Licensee Event Report (LER). This report fulfills the requirement for a written report within sixty (60) days of a reportable occurrence.

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

Sincerely, Georg/er SWR/swr lamrick resident ar Plant rogress x 10429 C 28461 57.3698

Enclosure:

Licensee Event Report ie~D~

cc (with enclosure):

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: Ms. Michelle P. Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission ATTN: Mr. Andrew Hon (Mail Stop OWFN 8G9A) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01131/2017 (02.2014)

Estimated burden per response to comply with this mandatory collection request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send comments regarding burden estimate to the FOIA, Privacy and Information Collections LER Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by LICENSEE EVENT REPORT (L )

intemet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and (See Page 2 for required number of Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently vafid OMB digits/characters for each block) control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Brunswick Steam Electric Plant (BSEP) Unit 1 05000325 1 OF 4

4. TITLE Fire Related Unanalyzed Condition that Could Impact Equipment Credited in Safe Shutdown Analysis

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED YA SEUNIL RVFACILITY NAME DOCKET NUMBER MONT DA YEAR YEAR SEQUENTIAL REV MONH DY YA NUMBER NO.

MONTH DAY YEAR BSEP Unit 2 05000324 FACILITY NAME DOCKET NUMBER 03 20 2014 2014-004 I0 05 16 2014 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

[] 20.2201(b)

El 20.2203(a)(3)(i)

[I 50.73(a)(2)(i)(C)

El 50.73(a)(2)(vii) 51 20.2201(d)

[

20.2203(a)(3)(ii)

LI 50.73(a)(2)(ii)(A)

El 50.73(a)(2)(viii)(A)

[]

20.2203(a)(1)

LI 20.2203(a)(4) 0 50.73(a)(2)(ii)(B)

El 50.73(a)(2)(viii)(B)

[] 20.2203(a)(2)(i)

D 50.36(c)(1)(i)(A)

[]

50.73(a)(2)(iii)

[]

50.73(a)(2)(ix)(A)

10. POWER LEVEL

[] 20.2203(a)(2)(ii)

[1 50.36(c)(1)(ii)(A)

[:] 50.73(a)(2)(iv)(A)

El 50.73(a)(2)(x)

[j 20.2203(a)(2)(iii)

LI 50.36(c)(2)

LI 50.73(a)(2)(v)(A)

[]

73.71(a)(4) 000 20.2203(a)(2)(iv)

LI 50.46(a)(3)(ii)

LI 50.73(a)(2)(v)(B)

L] 73.71(a)(5)

IZ 20.2203(a)(2)(v)

El 50.73(a)(2)(i)(A)

LI 50.73(a)(2)(v)(C)

LI OTHER 20.2203(a)(2)(vi)

LI 50.73(a)(2)(i)(B)

LI 50.73(a)(2)(v)(D)

Specify in Abstract below or in

Energy Industry Identification System (EIIS) codes are identified in the text as [XX].

Introduction

Initial Conditions On March 20, 2014, Unit 1 was in Mode 5 with the reactor vessel disassembled for refueling. Unit 2 was in Mode 1 at 100 percent of rated thermal power.

Reportability Criteria 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 NRC was initially notified of this event at 1134 Eastern Daylight Time (EDT) on March 20, 2014 (Event Number 49935).

Event Description

On March 20, 2014, in preparation for converting from 10 CFR 50 Appendix R to National Fire Protection Association (NFPA) Standard 805, a review of the BSEP Safe Shutdown Analysis (SSDA) was being performed. This review identified conditions that may not ensure a protected train of equipment remains available under certain postulated fire scenarios. The review identified the following fire scenarios on each unit in which a fire could result in disabling critical pieces of equipment.

To ensure net positive suction head (NPSH) requirements are met for Residual Heat Removal (RHR) [BO]

system pumps, Containment Over Pressurization (COP) must be maintained. This is accomplished by ensuring Reactor Building Closed Cooling Water (RBCCW) [CC] pumps remain off. Analysis determined that postulated fires in areas CB-23E, RB1-N, and RB1-S, located in the Main Control Room [NA] and in the Reactor Building [NG], could have affected the ability to secure the RBCCW pumps and keep them secured. This potentially affects Unit 1.

A postulated fire in areas TB1 and RB1-S, located in the Unit 1 Turbine Building [NM] 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 [NB]. This could disable the fans which are assumed operable in the SSDA. This potentially affects both units.

A postulated fire in area DG-07, located in the switchgear room for 480V bus E6 [ED], 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 480-volt bus E4 to 480-volt bus E2. This potentially affects both units.

A postulated fire in area DG-16E, located in the supply air plenum and exhaust fan area of the 50' 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 [EA]. This potentially affects both units.

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 [EK]. A postulated fire in this area could also could blow the fuse for the control circuitry for breakers that are fed from 480-volt 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.

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 Main Control Room and in the Reactor Building, could have affected the ability to secure the RBCCW pumps and keep them secured. This potentially affects Unit 2.

Event Cause

This event resulted from oversights in the SSDA performed in connection with the original implementation of 10 CFR 50, Appendix R. It was discovered as a result of performing a detailed revalidation of the original analysis. The postulated fire scenarios identified in this report were not previously identified because the analysis techniques, used when 10 CFR 50, Appendix R was enacted, were less detailed than those being employed in the review supporting transition to NFPA-805. Since the event is a historical condition originating in the 1980s, no root cause was identified.

Safety Assessment

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.

Corrective Actions

Any changes to the corrective actions and schedules noted below will be made in accordance with the site's corrective action program.

Hourly fire watches were already being performed in affected areas for reasons other than this event and are being maintained as an interim action.

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.

Previous Similar Events

Two events were reported in LER 1-2013-002, dated September 27, 2013, and LER 1-2011-002, dated December 8, 2011, in which the plant was found to be in an unanalyzed condition with respect to certain fire scenarios. Corrective actions previously applied could not have prevented this event because they were specific to the particular fire scenarios that had been identified. Moreover, the SSDA continued under detailed review in anticipation of implementing NFPA-805, and that review resulted in the discoveries described in this report.

Commitments

No regulatory commitments are contained in this report.