05000286/LER-2007-003, Re Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable Recirculation Pump Caused by a Potential Strong Pump-Weak Pump Interaction During a Small Break LOCA
| ML080160125 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/07/2008 |
| From: | Joseph E Pollock Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-07-150 LER 07-003-00 | |
| Download: ML080160125 (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 |
| 2862007003R00 - NRC Website | |
text
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 EntfJ1 Buchanan, N.Y. 10511-0249 Tel (914) 734-6700 J. E. Pollock Site Vice President January 7, 2008 Indian Point Unit No. 3 Docket No. 50-286 NL-07-150 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001
Subject:
Licensee Event Report # 2007-003-00, "Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable Recirculation Pump Caused by a Potential Strong Pump-Weak Pump Interaction During a Small Break LOCA"
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2007-003-00. The attached LER identifies an event where there was a Technical Specification prohibited condition that exceeded the Allowed Completion Time for a train of the Emergency Core Cooling System, which is reportable under 10 CFR 50.73(a)(2)(i)(B). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP3-2007-04212.
There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 734-6710.
Sincerely, j1 vweL J. E. Pollock Site Vice President Indian Point Energy Center cc:
Mr. Samuel J Collins, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 3 Mr. Paul Eddy, New York State Public Service Commission INPO Record Center
/J iz1(
Abstract
On November 8, 2007, during an NRC Component Design Basis Inspection, a condition was identified where a pump of the Emergency Core Cooling Internal Recirculation (IR)
System could be inoperable during operator response actions for certain Small Break Loss of Coolant Accidents (SBLOCAs).
The condition is due to a procedure requirement (Emergency Operating Procedures ES-1.3 and ES-I.4) to start a second IR pump resulting in a potential strong pump/weak pump interaction that could result in a less than acceptable flow rate through the weak pump causing it to become inoperable.
During past plant operation, this unknown condition resulted in exceeding the allowed outage time of Technical Specification 3.5.2 for an inoperable ECCS train.
The apparent cause of the condition was inadequate analysis during original plant design due to insufficient engineering rigor which failed to identify the strong pump/weak pump interaction of the recirculation pumps for SBLOCAs.
This condition was also determined to be applicable to unit 2.
Immediate corrective actions were to declare a train of IR inoperable and revise ES-l.3 and ES-I.4 to eliminate the requirement to start a second IR Pump for a SBLOCA.
The UFSAR will be revised and a review of other applicable safety related pumps will be performed for potential impact from strong pump/weak pump interactions and necessary actions taken.
The event had no effect on public health and safety.
(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)
Past Similar Events A review was performed of Licensee Event Reports (LERs) for the past three years for any events due to inadequate design analysis.
No LERs were identified that reported events based on this cause.
LER-2006-001 reported a reactor trip as a result of a main generator trip from a short in the differential protection circuit caused by wires that had an inadequate design in material application.
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 accidents.
- Also, the ECCS design is robust in that the RHRS provides a backup capability to the IRS
.and at least one RHR train would have been available during the time the 31 IRP pump was potentially inoperable.
As noted in FSAR Section 6.2.2, the RHRS provides a backup recirculation capability.
Under postulated accidents that are discussed in FSAR Section 14.2, the analyzed LOCAs assume a loss of offsitepower and a single failure disabling one ECCS train.
This analyzed condition would result in only one train of IRS thereby preventing the potential for a strong pump/weak pump interaction.
An assessment was performed to determine the impact of the condition on Core Damage Frequency (CDF).
The assessment of the IRP strong pump/weak pump interaction issue determined there would be a change in internal-events CDF of no more than 5E-7 per year.
That CDF impact is considered not significant (Green Band delta CDF/yr).
Because the issue only impacts recirculation, which generally has little impact on large early release frequency, the controlling concern was the impact on CDF.