05000325/LER-2010-002

LER-2010-002, JUN 2 3 2010
SERIAL: BSEP 10-0076 10 CFR 50.73
U. S. Nuclear Regulatory Commission
ATTN: Document Control Desk
Washington, DC 20555-0001
Subject: Brunswick Steam Electric Plant, Unit No. 1
Renewed Facility Operating License No. DPR-71
Docket No. 50-325
Licensee Event Report 1-2010-002
Ladies and 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 (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 Ms. Annette Pope, Supervisor
Licensing/Regulatory Programs, at (910) 457-2184.
Sincerely,
14,,o4ti 0_
Edward L. Wills, Jr.
Plant General Manager
Brunswick Steam Electric Plant
MAT/mat
Enclosure:
Licensee Event Report
Progress Energy Carolinas, Inc.
Brunswick Nuclear Plant
PO Box 10429
Southport, NC 28461
Document Control Desk
BSEP 10-0076 / Page 2
cc (with enclosure):
U. S. Nuclear Regulatory Commission, Region II
ATTN: Mr. Luis A. Reyes, Regional Administrator
245 Peachtree Center Avenue, NE, Suite 1200
Atlanta, GA 30303-1257
U. S. Nuclear Regulatory Commission
ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector
8470 River Road
Southport, NC 28461-8869
U. S. Nuclear Regulatory Commission (Electronic Copy Only)
ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A)
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Rockville, MD 20852-2738
Chair - North Carolina Utilities Commission
P.O. Box 29510
Raleigh, NC 27626-0510
NRC FORM 366U U.S. NUCLEAR REGULATORY COMMISSION
(9-2007)
LICENSEE EVENT REPORT (LER)
(See reverse for required number of
digits/characters for each block)
1. FACILITY NAME
Brunswick Steam Electric Plant (BSEP), Unit 1
4. TITLE
APPROVED BY OMB: NO. 3150-0104E EXPIRES: 08/31/2010
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2. DOCKET NUMBER 3. PAGE
05000325 1 of 5
Operation Prohibited by Technical Specifications - Reactor Protection System (RPS) Instrumentation

Brunswick Steam Electric Plant (Bsep), Unit 1
Event date:	04-25-2010
Report date:	06-23-2010
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
3252010002R00 - NRC Website

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

Introduction Initial Conditions At the time of the event, Unit 1 was in Mode 2, at 900 psig. Startup from the Unit 1, Cycle 18 refueling outage was in progress.

Reportability Criteria On April 24, 2010, at 0313 hours0.00362 days <br />0.0869 hours <br />5.175265e-4 weeks <br />1.190965e-4 months <br /> Eastern Daylight Time (EDT), Unit 1 entered Mode 2 during startup from the Unit 1, Cycle 18 refueling outage. On April 25, 2010, at 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />, with the unit in Mode 2 at 900 psig, pressure transmitter 1-B21-PT-N023B was declared inoperable due to reading downscale. This transmitter is part of the B1 channel Reactor Protection System [JC] instrumentation required for operability of Function 3, "Reactor Vessel Steam Dome Pressure - High," of Technical Specification (TS) 3.3.1.1, "Reactor Protection System (RPS) Instrumentation." Table 3.3.1.1-1, "Reactor Protection System Instrumentation," which requires two operable channels when in Modes 1 and 2. It was determined that the inoperability of pressure transmitter 1-B21-PT-N023B was due to valve 1-B21-IV-1384, "B21-PT- N023B Instrument Isolation Valve," being in the closed position and that this valve was closed prior to entering Mode 2.

This condition is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as operation prohibited by the plant's Technical Specifications. Unit 1 entered Mode 2 without satisfying the requirements of Limiting Condition for Operation (LCO) 3.0.4. Additionally, Unit 1 operated in Mode 2 for approximately 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> and 47 minutes prior to identifying the condition and entering Condition A of TS 3.3.1.1.

Event Description

On April 25, 2010, at 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />, pressure transmitter 1-B21-PT-N023B was declared inoperable due to reading downscale. The downscale reading was discovered during the normally scheduled performance of procedure 101-03.1, "Reactor Operator Daily Surveillance Report," for the Control Room back panel area.

Troubleshooting activities were initiated and it was determined that the inoperability of pressure transmitter 1-B21-PT-N023B was due to valve 1-B21-1V-1384 being in the closed position. The valve was repositioned and, after appropriate testing, pressure transmitter 1-B21-PT-N023B was declared operable at 0309 hours0.00358 days <br />0.0858 hours <br />5.109127e-4 weeks <br />1.175745e-4 months <br /> on April 26, 2010.

In addition, it was determined that valve 1-B21-PT-N045B-3, "B21-PT-N045B Instrument Isolation Valve," was also out of position. This rendered pressure transmitter B21-PT-N045B inoperable. This transmitter is part of the instrumentation required for operability of Function 2, "Reactor Vessel Steam Event Description (continued) Dome Pressure - High," of TS 3.3.4.1, "Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation." However, the Applicability for TS 3.3.4.1 is Mode 1 and the issue was identified and corrected prior to entry into Mode 1.

Event Cause The select cause of this event was the failure to effectively use concurrent verification during the performance of procedure OMST-EFCV18R, "EFCV Rx Inst Pen Sys Isol Vlv Func Test X53, X82, X49B-A." As a result, valve 1-B21-IV-1384 was left in the closed versus open position. A review was performed to confirm that performance of OMST-EFCV18R, between March 18 and March 20, 2010, was the only outage activity that could have resulted in the mispositioning of valve 1-B21-IV-1384. OMST-EFCV18R determines operability of several reactor instrument penetration Excess Flow Check Valves (EFCVs) in conformance with TS Surveillance Requirement 3.6.1.3.7 (i.e., verify a representative sample of reactor instrumentation line EFCVs actuate to the isolation position on an actual or simulated instrument line break signal).

The technicians involved with performing OMST- EFCV18R were individually interviewed regarding procedure compliance details. The concurrent process consisted of one individual implementing place- keeping and initialing of each step. The second individual-, in close proximity and able to concurrently read the step, performed valve manipulations. Gloves were used for potential contamination control while touching the valve handles. This individual would manipulate several valves and after completion of multiple procedure steps, remove the gloves and signoff the completed steps. The occurrence of an error during this evolution cannot be confirmed. However, it is believed that errors in place-keeping occurred that resulted in the step for restoration of the isolation valve for B21-PT-N023B (i.e., and B21-PT-N045B) being missed or otherwise incorrectly performed. This is substantiated by the fact that performance of OMST-EFCV18R was the only outage activity that could have resulted in the mispositioning of valve 1- B21-IV-1384.

A contributing cause to this event was a lack of sufficient barriers, in the Return to Service section of OMST-EFCV18R, to ensure proper system alignment at the completion of the surveillance test. The procedure requires the plant to be in Mode 5 and, as such, contains no means to functionally verify proper component operation upon completion of the test. Additionally, the procedure does not contain independent verification of critical steps to verify components are returned to service.

Safety Assessment The safety significance of this event is considered minimal. Pressure transmitter 1-B21-PT-N023B is part of the instrumentation required for operability of Function 3, "Reactor Vessel Steam Dome Pressure - Safety Assessment (continued) High," of TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," which generates a RPS actuation signal. The trip channels for this Function are as follows:

• Al - 1-B21-PT-N023A

• A2 - 1 -B21 -PT-N023 C

• B1 - 1-B21-PT-N023B

• B2 - 1 -B21 -PT-N023D In this case, the B1 channel was inoperable. Due to the one-out-of-two taken twice logic (i.e., Al or A2 and B1 or B2 will result in a RPS actuation), trip capability was maintained. Therefore, there was no loss of safety function.

Corrective Actions

The following corrective action to prevent recurrence has been identified.

• Operations procedure 001-01.02, "Operations Unit Organization and Operating Practices," currently contains proper concurrent verification methodology. Maintenance personnel will use the concurrent verification process described within 001-01.02 during performance of maintenance activities requiring concurrent verification. Implementation of this expectation, within the Maintenance organization, has been completed.

• OPS-NGGC-1303, "Independent Verification," is a fleet level procedure that provides instructions for performance of independent verification including concurrent and functional verification for each of Progress Energy's nuclear plants. This procedure will be revised to incorporate the concurrent verification methodology contained in 001-01.02. This procedure revision is currently scheduled to be completed by August 19, 2010.

Additional corrective actions include the following.

• Valve ,1-B21-IV-1384 was repositioned and, after appropriate testing, pressure transmitter 1-B21- PT-N023B was declared operable at 0309 hours0.00358 days <br />0.0858 hours <br />5.109127e-4 weeks <br />1.175745e-4 months <br /> on April 26, 2010.

• Procedure OMST-EFCV18R will be revised to include requirements for independent verification of critical steps. This procedure revision is currently scheduled to be completed by July 7, 2010.

• Other EFCV surveillance procedures will be revised, as necessary, to include requirements for independent verification of critical steps. The identified procedure revisions are currently scheduled to be completed by July 7, 2010.

Previous Similar Events

A review of LERs and corrective action program condition reports for the past three years identified no previous similar events where failure to effectively use the concurrent verification resulted in Unit 1 or Unit 2 operation prohibited by the plant's Technical Specification.

Commitments No regulatory commitments are contained in this report.