05000298/LER-2003-003, Failure to Evaluate Heat-up Rate Leads to Technical Specifications Prohibited Operation
| ML031400047 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/14/2003 |
| From: | Hutton J Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2003047 LER 03-003-00 | |
| Download: ML031400047 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2982003003R00 - NRC Website | |
text
NLS2003047 May 14, 2003 N
Nebraska Public Power District Always there when you need us U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Licensee Event Report No. 2003-003 Cooper Nuclear Station, NRC Docket 50-298, DPR-46 The subject Licensee Event Report is forwarded as an enclosure to this letter.
Sincerely, Plant Manager Irar Enclosure cc: Regional Administrator USNRC - Region IV Senior Project Manager USNRC - NRR Project Directorate IV-1 Senior Resident Inspector USNRC NPG Distribution INPO Records Center Records COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com
Abstract
On April 10, 2000, during performance of the ASME Class 1 System Leakage Test surveillance procedure for refuel outage RE-19, the Technical Specification (TS) limit for Reactor Coolant System (RCS) heat-up rate was exceeded in Reactor Recirculation (RR) [EIIS:AD] loop B. The failure to meet TS Surveillance Requirement acceptance criteria was not recognized, and the required evaluation to determine if the RCS is acceptable for operation was not performed prior to start up from the RE-19 refuel outage. On March 20, 2003, with Cooper Nuclear Station (CNS) in Mode 5 (Refueling) for refuel outage RE-21, a review of the surveillance procedure and past performance of the procedure was performed in support of a modification to replace temperature recorders. During this review the above condition was discovered.
This event was the result of inadequate procedural guidance for equalizing RCS temperatures in preparation for starting an idle RR pump, and evaluating available RCS temperature data.
Appropriate procedure revisions were completed by April 4, 2003, and the required evaluation was completed on April 12, 2003. Additional corrective actions are being evaluated and tracked by the CNS Corrective Action Program.
NRC FORM 366 (7-2001)
(If more space is required, use additional copies of (If more space is required, use additional copies of NRC Forrn 366A) discharge valve to induce a limited flow in the loop piping. It was not recognized that the combined effect of warming the idle loop and starting of the idle loop pump, which equalized the coolant temperature, exceeded the limits established as acceptance criteria for the surveillance requirement.
Technical Specifications require the heat-up rate to be less than or equal to 100 degrees Fahrenheit (F) when averaged over a one hour period. The actual heat-up rate for the idle RR loop was approximately 120 degrees F averaged over one hour.
The required evaluation was completed on April 12, 2003. The evaluation concluded that the RCS is acceptable for operation.
BASIS FOR REPORT This event is reportable in accordance with 1 OCFR50.73(a)(2)(i)(B) as any operation or condition prohibited by the plant's Technical Specifications."
CAUSE
Two causes of this event were identified. The procedures which direct the performance of the ASME Class 1 System Leakage Test, and the operation of the RR system lacked explicit guidance to monitor and control RR loop inlet temperature of the idle loop. The second cause is the lack of a methodical process to calculate the RCS heat-up rate, including the RR loops, resulted in not having the data required to identify an excessive heat-up rate and alert the operator.
SAFETY SIGNIFICANCE
Based on post-event analysis, the risk significance of exceeding the RR Loop B heat-up rate did not adversely effect, directly nor indirectly the CNS risk as described by the probabilistic risk assessment as established by the baseline reliability of components or equipment. The condition did not challenge a fuel, reactor coolant pressure, primary containment, or secondary containment boundary. The condition did not impact the plant's ability to safely shutdown or maintain the reactor in a safe shutdown condition.
In addition, analysis has established that the condition did not compromise the plant design requirements for safety functions or important to safety component functions.
CORRECTIVE ACTIONS
Immediate Actions
- 1.
ASME Class 1 System Leakage Test surveillance procedure was placed on Administrative Hold on March 20, 2003, to prevent recurrence of the event. Revisions to the surveillance procedure and system operating procedure were issued by April 4, 2003, with enhanced guidance for control of the idle RR Loop temperature.
- 2.
The Technical Specification required evaluation to demonstrate that the RCS is acceptable for operation was completed on April 12, 2003.
Long Term Actions A new surveillance procedure to monitor and calculate RCS heat-up and cool-down rates was issued on April 21, 2003.
(If more space is required, use additional copies of NRC Form 366A)
PREVIOUS EVENTS There have been no previous events associated with heat-up limits being exceeded during the ASME Class 1 System Leakage Test.
LER 94-015-02 reported two separate events where the reactor vessel bottom head drain, the vessel bottom head, and the vessel above skirt junction exceeded cool-down limits subsequent to plant trips on December 14, 1993 and March 2, 1994. Corrective actions associated with this event included operator requalification training and a Software Design Change Request to develop and implement software to calculate the required data enabling faster reactor recirculation pump recovery. A dynamic computer display in the control room provides the data to the operators.
Three previous reportable events attributed to procedural problems have occurred in recent years.
LER 2000-009-00, Failure to Recognize Entry Condition for Limiting Condition for Operation 3.4.5 Condition D Causes Plant Operation in Violation of Technical Specification, was attributed to inadequate procedural requirements for a timely independent verification of LCO entries. Corrective actions included Operations Management providing procedural intent clarification to Control Room personnel.
Station documents were revised to detail the requirement to complete an independent verification of LCO entries prior to beginning work.
LER 2001-003-00, Failure to Adequately Revise Procedures Resulted in Inadequate Fire Watches Under Certain Battery/Battery Charger Configurations and an Unanalyzed Condition, was the result of not appropriately incorporating requirements into procedures. Corrective actions included revision to appropriate plant procedures to require fire watches in the necessary fire zones whenever the "C" battery charger is substituting for the "B" charger.
LER 2002-001-01, Loss of High Pressure Coolant Injection Safety Function Due to Gland Seal Condenser High Level Annunciation, was attributed to the procedure change process which allowed inappropriate guidance to be incorporated into a procedure. Process improvements since 1993 have corrected this deficiency. The alarm response procedure was revised to remove the step to inhibit High Pressure Coolant Injection which caused this event.
I ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS Correspondence Number: NLS2003047 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the NL&S Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITTED DATE
COMMITMENT
OR OUTAGE Revisions to the surveillance procedure and system operating Complete procedure were issued with enhanced guidance for control of the idle RR Loop temperature.
Cooper Nuclear Station has implemented a new surveillance Complete procedure to monitor and calculate RCS heat-up and cool-down rates.
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