05000302/LER-2009-001
| Crystal River Unit 3 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| Initial Reporting | |
| ENS 44807 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation |
| 3022009001R00 - NRC Website | |
At 10:17, on January 27, 2009, Progress Energy Florida, Inc. (PEF), Crystal River Unit 3 (CR-3) was operating in MODE 1 (POWER OPERATION) at 100 percent RATED THERMAL POWER when the Control Room staff received multiple alarms and observed Reactor Coolant System (RCS) [AB] pressure rising toward the automatic reactor trip setpoint [JD]. The reactor [AC] was manually tripped prior to reaching the automatic reactor trip setpoint. Emergency Operating procedure EOP-2, "Vital System Status Verification," was entered before eventually transitioning to EOP-10, "Post-Trip Stabilization.
Prior to this event, two experienced relay technicians from the Progress Energy Florida Substation Maintenance Department were tasked with performing Work Order Task 1072772-01 utilizing Preventive Maintenance (PM) procedure PM-290, "Calibration of Switchboard Meters and Transducers." While setting up to perform calibration on the 'A' 4160 volt (V) Unit Bus [EB, BU] indication in the Main Control Board, a relay technician was verifying voltage on the 'A' 4160V Unit Bus potential transformer (PT) [EB, XPT] prior to opening the links and connecting the Doble Relay Test Set. This is accomplished by using a Fluke Multimeter with blue and yellow test leads. (The Fluke Multimeter blue and yellow test leads were a result of the corrective actions established by Nuclear Condition Report (NCR) 133661. At all other Progress Energy plants, the Fluke Multimeter test leads are red/black.) The black and red test leads from the Doble Relay Test Set were incorrectly connected in place of the Fluke Multimeter blue and yellow test leads during performance of a live-dead-live voltage check. This resulted in a path to ground that blew the 'A' and 'B' Secondary Side PT fuses [EB, FU] on the 'A' 4160V Unit Bus and a loss of voltage indication to the 'A' 4160V Unit Bus loads.
This produced the actuation of undervoltage motor protection relays [EB, 27], tripping of the 4160V loads attached to the 'A' 4160V Unit Bus (e.g., Circulating Water System (CW) [KE] pumps CWP-1A/1C [KE, P], Main Feedwater System (FW) [SJ] pump FWP-1A [SJ, P], Condensate System (CD) [SG] pump CDP-1A [SG, P], Secondary Services Closed Cycle Cooling System (SC) [KB] pump SCP-1A [KB, P], etc). Nuclear Services and Decay Heat Raw Water System (RW) [KI] pump RWP-2B [KI, P] and SCP-1B automatically started per design.
Upon initiation of the manual reactor trip, the main turbine [TA] automatically tripped and the 'A' 4160V Unit Bus transferred from the Unit Auxiliary Transformer [EB, XMFR] to the Startup Transformer per design. The 'A' 4160 Unit Bus and the 'A' 480V Bus remained energized.
At 11:03, on January 27, 2009, the MODE 3 (HOT STANDBY) lineup was established. At 12:49 on January 27, 2009, EOP-10 was exited.
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-2B suction line relief valve FWV-16 [SJ, RV] lifted and did not reseat.
_NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER The 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 manual reactor trip is bounded by the Final Safety Analysis Report accident analysis.
Based on the above discussion, PEF concludes that the RPS performed as designed and did not represent a reduction in the public health and safety. Since no loss of safety function occurred, this event does not meet the Nuclear Energy Institute (NEI) definition of a Safety System Functional Failure (NEI 99-02, Revision 2).
CAUSE
Two causes have been identified for the need to manually trip the reactor. The first cause was improper use of human performance tools which led to the connection of the incorrect test leads to the 'A' 4160V Unit Bus. The three causal factors for this human error were: 1) self-checking not applied to ensure correct intended action; 2) failure to effectively use peer checking; and, 3) procedure use and adherence failure (i.e., performance of the voltage check outside of the work order guidance).
The second cause was inadequate closure of corrective actions for a similar event that occurred at CR-3 in 2004 (Priority 2 NCR 133661). The CR-3 Plant Nuclear Safety Committee (PNSC) established an action to: "Add a corrective action to complete a risk assessment for this type of work (calibrating volt meters on the main control board), whether it should be completed on-line vs. off-line." The response to this corrective action was not adequate and was not reviewed by the PNSC or the PNSC Chairman. This resulted in the failure to move relay activities that could result in a plant transient from on-line to outage.
�NRC FORM 366A (9-2007) PRINTED ON RECYCLED PAPER Two of three fuses are needed to blow in order to result in actuation of undervoltage relays and loss of voltage indication to the 'A' 4160V Unit Bus loads.
ATTACHMENTS
Attachment 1 — Abbreviations, Definitions, and Acronyms Attachment 2 — List of Commitments _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 0