05000269/LER-2010-002, Regarding Manual Reactor Trip Due to 1A1 and 1A2 Reactor Coolant Pump High Vibration Indication
| ML102860133 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/06/2010 |
| From: | Gillespie T Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 10-002-00 | |
| Download: ML102860133 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 2692010002R00 - NRC Website | |
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T. PRESTON GILLESPIE, Jr.
Duke Vice President [wEnergy Oconee Nuclear Station Duke Energy ON01 VP / 7800 Rochester Hwy.
Seneca, SC 29672 864-873-4478 864-873-4208 fax T. Gillespie@duke-energy. corn October, 6, 2010 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555
Subject:
Oconee Nuclear Station Docket Nos. 50-269 Licensee Event Report 269/2010-02' Revision 0 Problem Investigation Process No.: 0-10-6174 Gentlemen:
Pursuant to 10 CFR 50.73 Sections (a)(1) and (d), attached is Licensee Event Report 269/2010-02, Revision 0, regarding a Unit 1 manual reactor trip due to incorrect indications of increasing reactor coolant pump high vibration.
This report is being submitted in accordance with 10 CFR 50.73 (a)(2)(iv)(A).
This event is considered to be of no significance with respect to the health and safety of the public.
There are no regulatory commitments contained in this report.
Cause analysis for this event is not yet complete. Results will be provided in a supplement to this report.
Any questions regarding the content of this report should be directed to Corey Gray, Regulatory Compliance Group at 864-873-6325.
Sincerely, T. Preston Gillespie Jr.,
Vice President Oconee Nuclear Station Attachment www.duke-energy.comr
Document Control Desk October 6, 2010 Page 2 cc:
Mr. Luis Reyes Administrator, Region I1 U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Mr. John Stang Project Manager U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C. 20555 Mr. Andrew Sabisch NRC Senior Resident Inspector Oconee Nuclear Station INPO (Word File via E-mail)
Abstract
On August 7, 2010, at 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br />, while operating at 100%, Oconee Nuclear Station (ONS) Unit 1 received statalarm 1 SA-9-D2 (Reactor Coolant Pump Vibration High). Operators entered the applicable response procedure that resulted in Unit 1 reducing power. Vibration readings continued to increase. At 1451 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.521055e-4 months <br />, with power at 17%, Control Room Operators manually tripped Unit 1 when the indicated vibration readings reached procedural limits. The post trip response was normal, with all major operating parameters remaining within expected limits. Operators took appropriate action to stabilize the unit in Mode 3 (hot standby). While the operators were controlling Pressurizer and Letdown Storage Tank Level (per procedure) the standby 1 B High Pressure Injection (HPI) pump auto started due to low Reactor Coolant Pump Seal Flow.
Investigation revealed that a false high Reactor Coolant Pump (RCP) vibration indication occurred due to a failed power supply within the control module.
Corrective Actions included replacement of the Unit 1 Control module (500-1M22) and positioning the RCPs vibration equipment Power Supply Selector switches to "Auto".
This event is considered to have no significance with respect to the health and safety of the public.
NRC FORM 366 (7-2001)
(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 Form 366A) the trip, no safety limits were challenged and the Unit was placed in Mode 3 without complications.
Consequently this event was not considered safety significant.
No safety limits were challenged and shutdown was as expected for a manual trip scenario. Based on analysis this event has a low impact on core damage risk.
ADDITIONAL INFORMATION
A determination of whether there were previous similar events is pending completion of the Root Cause evaluation and a review of the Corrective Action Program data base. A summary of results will be submitted in a supplement to this report.
Energy Industry Identification System (EIIS) codes are identified in the text within brackets [.
There were no releases of radioactive materials, radiation exposures or personnel injuries associated with this event.
This event has been reported to the Equipment Performance and Information Exchange (EPIX) program. The failed component was a model 500-IM22 vibration monitor control module made by Indikon Company Inc.