05000324/LER-2005-004, Re Loss of Electrical Power to Emergency Bus E1

Re Loss of Electrical Power to Emergency Bus E1

ML052010044
Person / Time
Site:	Brunswick
Issue date:	07/11/2005
From:	Hinds D Progress Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 05-0075 LER 05-004-00
Download: ML052010044 (12)

LER-2005-004, Re Loss of Electrical Power to Emergency Bus E1

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2) 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)()(B)
3242005004R00 - NRC Website

text

Progress Energy July 11, 2005 SERIAL: BSEP 05-0075 10 CFR 50.73 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. I and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-72 Licensee Event Report 1-2005-004 Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.73, Carolina Power

& Light Company, now doing business as Progress Energy Carolinas, Inc., submits the enclosed Licensee Event Report. 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. Edward T. O'Neil, Manager-Support Services, at (910) 457-3512.

Sincerely, David H. Hinds Plant General Manager Brunswick Steam Electric Plant CRE/cre Enclosure: Licensee Event Report Progress Energy Carolinas, Inc.

Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461

Document Control Desk BSEP 05-0075 / Page 2 cc (with enclosure):

U. S. Nuclear Regulatory Commission, Region II AT'TN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATIN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission ATIN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) (Electronic Copy Only) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-051

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150.0104 EXPIRES 06/30/2007 (6-2004)

Essmated buden per msporse to cuAy tv tis w

endarty Wormrtion coneecon request 50 hocurs. Repoirted lessons learned are kicorporate kW to kwie iceng procss "r fed back lo LICENSEE EVENT REPORT (LER) h~stry.Sendc=nw regaring bbrden smWeDterd Sieb (See reverse for required number of literriet e-fl to -dcc letsanrcgo, and to the Desk Offir, Office d hWnTabonw4ad digits/characters for each block)

Reqratory Affairs, NEOB-10202 (3150-0104) OffiTe of Managrfl and Budget Wastiinrtm DC 20503. I a mman used to kirpose an nforrmation ociecdtn does not dspiay a cmrarty valid OMB c o nturer, tie NRC mnay nd cxWr or sponsor.

and a person is id reired to respond lo ve I rmbon cdlecuort

3. PAGE Brunswick Steam Electric Plant (BSEP), Unit 1 05000325 1 OF 10

4. TITLE Loss of Electrical Power to Emergency Bus El

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILMES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTHDAYYEAR FACILUTY NAME DOCKETNUMBER MOYA ER NUMBER NO MOT A

ERBSEP, Unit 2 05000324 05 l12 205 205 -

4 07 11 205 FACILITY NAME DOCKET NUMBER 05~205 005--

04 0~~j005_________

05000

9. OPERATING MODE

11. THIS REPORTIS SUBMrIrED PURSUANTTO THE REQUIREMENTS OF 10 CFR §: (Checkoneormore) 1 20.2201 (b)

D 20.2203(a)(3)(i) 0 50.73(a)(2)(i (C) 0 50.73(a)(2)(vii) o 20.2201 (d) a 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A) o 20.2203(a)(1) 0 20.2203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B)

El 20.2203(a)(2)(i)

El 50.36(c)(1)(i)(A)

[I 50.73(a)(2)(iii) 0 50.73(a)(2)(ix)(A)

10. POWER LEVEL El 20.2203(a)(2)(ii) 0 50.36(c)(1)(ii)(A)

[

50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) 097 0 20.2203(a)(2)(iii) 0 50.36(c)(2)

El 50.73(a)(2)(v)(A) 0 73.71 (a)(4) o 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii)

E 50.73(a)(2)(v)(B)

El 73.71 (a)(5) o 20.2203(a)(2)(v)

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

O 50.73(a)(2)(v)(C)

OTHER 0

5.7(a)2)i)IB)Spcf In Abstract below 0

20.2203(a)(2)(vi)

ED 50.73(a)(2)()(B) 50.73(a)(2)(v)(D) or in hc (If more space is required, use additional copies of (If more space Is required, use additional copies of (If more space Is required, use additional copies of (If more space Is required, use additional copies of (If more space Is required, use additional copies of (If more space Is required, use additional copies of NRC Forn 366A) (17)RCS Leakage Detection Instrumentation Inoperability (continued) granting the enforcement discretion. On May 17, 2005, PEC submitted a letter, SERIAL:

BSEP 05-0060, Subject: Request for License Amendment, Technical Specification 3.4.5, "RCS Leakage Detection Instrumentation," for Units 1 and 2. On June 28, 2005, the NRC issued a letter, Subject:

BSEP Units 1 and 2 - Issuance of Amendment on Technical Specification 3.4.5, Reactor Coolant System Leakage Detection Instrumentation (TAC NOS. MC7216 and MC7217), revising the TS to replace Required Action D.1 (i.e., immediately enter LCO 3.0.3) with the requirement to be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> if all required leakage detection systems are inoperable.

Control Room AC and CREV Systems Loss of Safety Function The broken ring lug was replaced with a heavy duty ring lug, the 2A CB Air Compressor was started, and the Control Room AC and CREV systems were returned to service. The 2B CB Air Compressor control wiring was subsequently inspected satisfactorily with no similar conditions existing. Procedures for performing breaker maintenance were reviewed and determined to contain proper guidance regarding inspection and termination of control wiring. Operating experience is being developed to make maintenance personnel aware of this occurrence.

SAFETY ASSESSMENT

The safety significance of this event is considered to be minimal. Operation of the plants in response to the system/equipment inoperabilities was in accordance with TS LCO Required Action Completion Times, with the exception of Unit 1 operation with regard to LCO 3.4.5. An NRC NOED was requested and received to delay the Unit 1 shutdown required by LCO 3.4.5. Discussion of the major system/equipment impacts follows.

Loss of Electrical Power to Bus El Electrical power distribution subsystems are required to be operable in Modes 1, 2, and 3 to ensure that acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of anticipated operational occurrences or abnormal transients. The subsystems are also required to be operable to ensure that adequate core cooling is provided, and containment operability and other vital functions are maintained in the event of a postulated design basis accident (DBA). With one or more required AC buses in one division inoperable, the remaining AC electrical power distribution subsystems are capable of supporting the minimum safety functions to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure. The overall reliability is reduced; however, because a single failure in the remaining AC electrical power subsystems could result in the minimum required engineered safety feature functions not being supported. For this event, bus El and its supported buses were de-energized for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and 29 minutes, and were considered to be inoperable for approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> and four minutes. During that time, the remaining AC electrical subsystems in the other division and their supported systems/subsystems remained operable to safely shut down the reactors and maintain them in a safe shut down condition, if needed. Maintenance and operational activities were restricted during the event to minimize challenges to the operators and equipment.

(if more space Is required, use additional copies of (If more space Is required, use additional copies of NRC Form 366A) (17)RCS Leakage Detection Instrumentation Inoperability (continued)

Unit 1 operated for approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and 29 minutes with all required RCS Leakage Detection Instrumentation inoperable. Prior to the event, Unit 1 unidentified RCS leakage was well below the LCO 3.4.4 operational leakage rate limit. Increased monitoring by operators during the event did not detect an RCS leakage increase, and when the RCS Leakage Detection Instrumentation was returned to service, RCS leakage was confirmed to be at pre-event rates.

Control Room AC and CREV Systems Loss of Safety Function The Control Room AC system is designed to provide a controlled environment under both normal and accident conditions. Two of the three subsystems provide the required temperature control to maintain a habitable environment and ensure the operability of components in the control room. Each subsystem consists of a heating coil, a cooling coil, a supply fan, a compressor-condenser unit, ductwork, dampers, and instrumentation and controls to provide for control room temperature control. In Modes 1, 2, and 3, the Control Room AC system must be operable to ensure that control room temperature will not exceed equipment operability limits following control room isolation.

The CREV system provides a radiologically controlled environment from which the units can be operated following a DBA. The system is a standby system that is common to both Units 1 and 2, parts of which also operate during normal unit operations to maintain the control room environment. The safety function of the CREV system is the radiation protection portion of the radiation/smoke protection mode and includes two redundant high efficiency air filtration subsystems for emergency treatment of recirculated air or outside supply air. Each subsystem consists of a high efficiency particulate air filter, an activated charcoal adsorber bank, an emergency recirculation fan, and the associated ductwork and dampers. In Modes 1, 2, and 3, the CREV system must be operable to control operator exposure during and following a DBA, since the DBA could lead to a fission product release.

Both Units 1 and 2 operated without the CB Air Compressors providing support for the Control Room AC and CREV systems operation for approximately one hour and 16 minutes. Plant staff took immediate actions to investigate/correct the problem and return the systems to service. For the brief time the Control Room AC and CREV systems were inoperable, performance of plant personnel and equipment were not adversely affected.

PREVIOUS SIMILAR EVENTS

A review of LERs and corrective action program condition reports for the past three years has not identified any previous similar events.

COMMITMENTS

No regulatory commitments are contained in this report. Those actions discussed in this submittal will be implemented in accordance with corrective action program requirements.