05000302/LER-2008-003
| Crystal River | |
| Event date: | 08-24-2008 |
|---|---|
| Report date: | 12-10-2008 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3022008003R01 - NRC Website | |
At 15:53 on August 24, 2008, Progress Energy Florida, Inc. (PEF), Crystal River Unit 3 (CR-3) was operating in MODE 1 (POWER OPERATION) at approximately 60 percent RATED THERMAL POWER, when the reactor was manually tripped due to instability of several key plant parameters, including reactor power, Main Feedwater (FW) System [SJ] flow and Once-Through Steam Generator (OTSG) [SB, SG] level.
At 15:37 on August 24, 2008, the Condensate Pump CDP-1A [SD, P] magnetic coupling (Electric Machinery, Model MSD4404V) [SD, CPLG] became uncoupled and resulted in numerous alarms in the Main Control Room. The alarms were a result of, but not a direct indication of, the failure. This resulted in distractions and significant lowering of Condensate System [SD] flow. Lowering Deaerator (FWHE-1) [SD, TK] level was not immediately diagnosed. In response to a lowering FWHE-1 level and recognition that CDP-1A amperage was well below normal, efforts to re-couple CDP-1A were directed, concurrent with entry into Abnormal Operating Procedure AP-510, "Rapid Power Reduction." Reactor power was reduced from 100 percent RATED THERMAL POWER and stabilized at approximately 62 percent RATED THERMAL POWER with the FWHE-1 level recovering.
At this time, FW block valve FWV-29 [SJ, ISV] to the "B" OTSG began to close, lowering the "B" OTSG leVel. FW block valve FWV-30 to the "A" OTSG remained full open. Oscillations were observed to be occurring on Main FW pump FWP-2B [SJ, P] flow between 0-100 percent demand while FWP-2A flow was pegged high. In response to the high FW System [SJ] flow condition on the "A" OTSG, FWV-30 was placed in manual and closed. FW System flow oscillations were still occurring, along with resultant reactor power oscillations, so the decision was made to manually trip the reactor. A turbine trip occurred simultaneously. Following completion of post-trip actions, FWP-2B oscillations continued. FW cross-connect valve FWV-28 was opened and FWP-2B was manually-tripped.
No structures, systems or components were inoperable at the start of the event that contributed to the event. No other pertinent maintenance or surveillance activities were in progress. Plant protection and non-protection systems operated normally during the manual reactor trip, with the exception of the following:
FWP-2A locked in at 100% demand during the power reduction causing control issues and overfeed of the "A" OTSG. The FWP-2A Woodward 505 digital governor controller [SJ, 65] interpreted a transient condition as a control failure due to a circuit design feature and locked in at the last good signal.
At 18:03 on August 24, 2008, a 4-hour notification to the NRC Operations Center (Event Number 44438) was made in accordance with 10 CFR 50.72(b)(2)(iv)(B) for manual actuation of the Reactor Protection System (RPS) [JC]. An update was provided at 19:21 on August 24, 2008.
This condition is being reported as a 60-day Licensee Event Report under 10 CFR 50.73(a)(2)(iv)(A) for manual actuation of the RPS.
Manual actuation of the RPS was initiated to shut down the reactor and maintain adequate OTSG levels. Upon initiation of the manual reactor trip, the RPS responded as expected, control rods fully inserted and safety systems functioned as required. No challenges to the RPS setpoints were identified. No Emergency Feedwater Initiation and Control System [BA] actuation occurred or was expected. Both OTSGs were fed by the FW System throughout this event.
This event did not result in the release of radioactive material. No design safety limits were exceeded and no fission product barriers or components were damaged as a result. The loss of Feedwater is an event analyzed and bounded by the Final Safety Analysis Report accident analysis.
Based on the above discussion, PEF concludes that manual actuation of the RPS did not represent a reduction in the public health and safety.
This event is not reportable under 10 CFR 50.73(a)(2)(v) and does not represent a condition that would have prevented the fulfillment of a safety function. Therefore, this event does not meet the Nuclear Energy Institute (NEI) definition of a Safety System Functional Failure (Reference: NEI 99-02, Revision 5).
CAUSE
The root cause for this event was original plant design failing to provide adequate alarms to the operating crew to promptly identify Condensate Pump failures. No Condensate Pump uncoupled alarm was or should have been received in the Control Room following the loss of CDP-1A. These alarms actuate when a low magnetic coupling supply voltage is sensed. Since these alarms originate in the Condensate Pump controllers, a failure of the Condensate Pump magnetic coupling itself, as occurred in this event, will not result in an alarm.
Corrective Actions
1. A Just-In-Time training package was developed for this event and operating crews were trained prior to assuming watch standing duties with the reactor critical. Operator selected alarms were established to alert the operating crews to Condensate Pump failures as an interim compensatory measure.
2. The CDP-1A motor and clutch assembly was replaced with a refurbished spare under Work Order 1406343. (Nuclear Condition Report (NCR) 294686) 3. The FWP-2A Woodward 505 digital governor controller was modified by Engineering Change 71126 (Work Order 1407302) to limit the signal range received from the Integrated Control System, bypassing the lockout feature. (NCR 293609) 4. Design options are being evaluated in the CR-3 Corrective Action Program to alert the operating crew of Condensate Pump failures. (NCR 293080-04) events revealed a problem with the original plant design associated with Condensate Pump loss indication in the Control Room. No previous similar events have been reported to the NRC.
ATTACHMENTS
Attachment 1 — Abbreviations, Definitions, and Acronyms Attachment 2 — List of Commitments N NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER The following table identifies those actions committed by PEF in this document. Any other actions discussed in the submittal represent intended or planned actions by PEF. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Supervisor, Licensing and Regulatory Programs of any questions regarding this document or any associated regulatory commitments.
COMMITMENT DUE DATE
No new regulatory commitments are contained in this submittal. N/A �