05000247/LER-2007-001, For Indian Point, Unit 2, Regarding Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable Residual Heat Removal Pump Due to an Electrical Supply Breaker Failure
| ML070650407 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/28/2007 |
| From: | Dacimo F Entergy Nuclear Operations |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| NL-07-013 LER 07-001-00 | |
| Download: ML070650407 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(ix) |
| 2472007001R00 - NRC Website | |
text
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 nBuchanan, N.Y. 10511-0249 Entfffl(Tel (914) 734-6700 Fred Dacimo Site Vice President Administration February 28, 2007 Indian Point Unit No. 2 Docket No. 50-247 NL-07-013 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001
Subject:
Licensee Event Report # 2007-001-00, "Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable Residual Heat Removal Pump Due to an Electrical Supply Breaker Failure"
Dear Sir:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2007-001-00. The enclosed LER identifies an 1
event where the plant was operated in a condition prohibited by Technical Specifications, which is reportable under 10 CFR 50.73(a)(2)(i)(B). This condition has been recorded in the Entergy Corrective Action Program as Condition Report CR-IP2-2007-00013.
There are no commitments contained in this letter. Should you or your staff have any questions regarding this matter, please contact Mr. Patric W. Conroy, Manager, Licensing, Indian Point Energy Center at (914) 734-6668.
Sincerely, red R. D~acimo ite Vice President Indian Point Energy Center
Docket No. 50-247 NL-07-013 Page 2 of 2 Attachment: LER-2007-001 -00 cc:
Mr. Samuel J. Collins Regional Administrator - Region I U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident Inspector's Office Resident Inspector Indian Point Unit 2 Mr. Paul Eddy State of New York Public Service Commission INPO Record Center
Abstract
On January 2, 2007, during quarterly surveillance testing of the 21 Residual Heat Removal pump (RHRP),
there was a failure of the pump to start on demand.
Investigation into the cause of the condition identified that the pump electrical supply breaker control power fuse had blown due to a breaker malfunction.
The supply breaker was a DB-50, 480 volt AC breaker.
The apparent cause of the breaker failure to close was a mispositioned inertia latch.
The lack of inertia latch reset caused the breaker closing coil to remain energized following breaker closure initiation until the control fuses actuated.
The inertia latch is designed to reset following a breaker trip to allow for future breaker closures.
The binding inertia latch was not resetting as required resulting in the failure of the breaker to close.
A contributing cause was the presence of a residue found on the inertia latch and pivot pin mating surfaces.
Corrective actions included breaker cleaning, re-lubrication, testing and return to service.
Subsequently, the 21 RHRP breaker was replaced with a refurbished breaker.
An examination of the breaker residue material will be performed by the breaker manufacturer and a report provided.
Engineering will review the report and any necessary corrective actions will be performed.
Although Engineering believes the breaker had freedom of movement, the breaker maintenance and modification procedures were enhanced to assure freedom of movement of the inertia latch.
An inspection of the remaining DB-50 breakers except closed breakers was performed.
All inspected breakers were found to be in working order.
The event had no effect on public health and safety.
NRC FORM 366 (6-2004)
(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) (17)
During the time the 21 RHR breaker was inoperable, the 22 RHR pump was operable and available to perform the safety function.
A review of support systems for the 22 RHR pump identified that the 23 Emergency Diesel Generator (EDG),
which provides emergency power for the 22 RHR pump, had been declared inoperable on October 29, 2006, as a result of EDG service water cooling water piping repairs.
During that time, offsite power remained available and would have supplied normal power for the 22 RHR pump.
At no time during the failed 21 RHR breaker condition was the RHR system unable to perform its safety function.
In accordance with reporting guidance in NUREG-1022, an additional random single failure need not be assumed in that system during the condition.
Therefore, there was no safety system functional failure of the RHR system reportable under 10 CFR 50.73(a) (2) (v).
Review of the condition for reporting under 10 CFR 50.73(a) (2) (ix),"Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems," determined the event is not reportable under this criterion.
Although other DB-50 breakers could have the condition and be in other safety trains that were not inspected, engineering judged that this failure was an isolated case and that the breakers, which now have new latches, have an excellent performance record and no record of failure.
Engineering judgment, as allowed by the guidelines of NUREG-1022, concluded that there is reasonable expectation that the safety functions of potentially affected systems could be fulfilled.
PAST SIMILAR EVENTS A review was performed of Licensee Event Reports (LERs) for the past three years for any events that involved DB-50 breaker failures that resulted in exceeding TS allowed completion times.
No LERs were identified that reported breaker failures.
SAFETY SIGNIFICANCE
This event had no effect on the health and safety of the public.
There were no actual safety consequences for the event because there were no events during the time the 21 RHR pump was inoperable due to its failed breaker condition.
In addition, the redundant 22 RHR pump was available to perform the safety function.
There were no significant potential safety consequences of this event under reasonable and credible alternative conditions.
The condition identified on January 2, 2007, would have prevented the 21 RHR breaker from closing since the last time it was successfully tested on October 9, 2006 (exposure time of 86 days).
A risk assessment was performed for this condition with the following results:
The Core Damage Frequency (CDF) was determined to be 1.812E-5 per year.
Given a baseline CDF of 1.787E-5 per year, the condition represents an incremental CDF (ICDF) of 2.50E-7 per year.
The Incremental Core Damage Probability (ICDP) based on the exposure time of 86 days (86 days/365 days per year) was determined to be 5.89E-8.
The ICDP determined for this event is below the value considered risk significant.