05000324/LER-2013-002, From Brunswick Steam Electric Plant, Unit 1, Fire Related Unanalyzed Condition That Could Impact Equipment Credited in Safe Shutdown Analysis
| ML13283A015 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/27/2013 |
| From: | Hamrick G Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BSEP 13-0106 LER 13-002-00 | |
| Download: ML13283A015 (6) | |
| 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 |
| 3242013002R00 - NRC Website | |
text
DUKE ENERGY.
George T. Hamrick Vice President Brunswick Nuclear Plant Duke Energy Progress P.O. Box 10429 Southport, NC 28461 o: 910.457.3698 10 CFR 50.73 SEP 2 7 2013 Serial: BSEP 13-0106 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket No. 50-325 and 50-324 Licensee Event Report 1-2013-002 In accordance with the Code of Federal Regulations, Title 10, Part 50.73, Duke Energy Progress, Inc., formerly known as Carolina Power & Light Company (CP&L) submits the enclosed Licensee Event Report (LER). This report fulfills the requirement for a written report within sixty (60) days of a reportable occurrence. Also, this report contains no regulatory
commitments
Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.
Sincerely, George Hamrick GTH/swr
Enclosure:
Licensee Event Report
U.S. Nuclear Regulatory Commission Page 2 of 2 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. Christopher Gratton (Mail Stop OWFN 8G9A) 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: 10/31/2013 (10-2010)
, 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 MONTH DAY YEAR YEAR SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER NUMBER NO.
MONTH DAY YEAR BSEP, Unit2 05000324 FACILITY NAME DOCKET NUMBER 07 29 2013 2013-002-00 09 27 2013
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
[L1 20.2201(b)
W 20.2203(a)(3)(i)
[
50.73(a)(2)(i)(C)
L] 50.73(a)(2)(vii) 1 E 20.2201(d)
[
20.2203(a)(3)(ii)
[
50.73(a)(2)(ii)(A)
EW 50.73(a)(2)(viii)(A)
[W 20.2203(a)(1)
ED 20.2203(a)(4) 0 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
El 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
[] 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL E] 20.2203(a)(2)(ii)
[1 50.36(c)(1)(ii)(A)
F-50.73(a)(2)(iv)(A)
[L1 50.73(a)(2)(x)
EL 20.2203(a)(2)(iii)
El 50.36(c)(2)
[] 50.73(a)(2)(v)(A)
L] 73.71(a)(4) 100 El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
[] 73.71 (a)(5)
EL 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
LI 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
[L 50.73(a)(2)(i)(B)
LI 50.73(a)(2)(v)(D)
Specify in Abstract below or in direct response personnel to open the breakers. The relevant procedures did not contain these steps.
As described above on Unit 1, this would interfere with the ability of the emergency diesel generators to power certain safety related loads. These loads are:
Residual Heat Removal system pump 2B Nuclear Service Water [BI] system pump 2B 480-volt circuit supporting motor control centers for Train B safe shutdown components
- 3. In fire area RB2-S located in the Unit 2 Reactor Building, the relevant fire response procedure should have instructed personnel to de-energize a control power breaker and depress the manual trip pushbutton for the 2A Control Rod Drive [AA] system pump power breaker. The absence of these steps creates the possibility that DC control power could be lost from the breaker while it was still closed.
This, in turn, would prevent the emergency diesel generator load sequencing timer from being reset, interfering with the ability to re-power certain safety-related loads from the emergency busses if offsite power were lost. These loads are:
Residual Heat Removal system pump 2A Nuclear Service Water system pump 2A 480-volt circuit supporting motor control centers for Train A safe shutdown components
- 4. The revalidation of the safe shutdown analysis also revealed issues with the assumptions about air flow to certain items of critical equipment. For postulated fires in fire areas DG-07 and DG-08E, ventilation dampers were susceptible to failing in the closed position, depriving safety-related electrical transformers and switchgear of cooling air. Fire response procedures should have directed personnel to secure certain doors in the open position in order to ensure a flowpath for cooling.
Event Cause
This event occurred as a result of oversights in the safe shutdown analysis 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. 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 were already ongoing in some of the areas prior to the time of discovery for reasons not related to this event. Other fire watches were established as a result of the discoveries made in 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.
NRC FORM 366 (10-2010)
Corrective Actions
Hourly fire watches were established or continued in all affected areas as an interim action.
Procedures 1ASSD-04, 1ASSD-05, 2ASSD-03, 2ASSD-05, and 2ASSD-06 will be revised to prescribe the required actions for mitigating the effects of a fire in the affected areas.
Previous Similar Events
A review of LERs and corrective action program condition reports identified a previous similar event. LER 1-2011-02 (ADAMS accession number ML 1355A1 55) reported the discovery that the plant was susceptible to spurious actuations of critical components related to several different fire areas. Corrective actions implemented included establishing fire watches in affected areas and revising procedures to address the identified vulnerabilities. The actions were specific to the discoveries made and thus would not have prevented this most recent event.
Commitments
No regulatory commitments are contained in this report.
NRC FORM 366 (10-2010)