05000311/LER-2012-001
| Salem Generating Station - Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3112012001R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
Westinghouse — Pressurized Water Reactor (PWR/4) * Energy Industry Identification System {EIIS} codes and component function identifier codes appear as {SS/CCC}
IDENTIFICATION OF OCCURRENCE
Event Date: March 23, 2012 Discovery Date: March 23, 2012
CONDITIONS PRIOR TO OCCURRENCE
Salem Unit 2 was in Operational Mode 1 at 100% power with the 22 Station Power Transformer (SPT) tagged out of service for scheduled maintenance. No other structures, systems or components were inoperable at the time of the discovery that contributed to the event.
DESCRIPTION OF OCCURRENCE
On March 23rd, 2012 Salem Unit 2 was in Mode 1, operating at 100% power and steady state conditions with the 22 SPT tagged out of service for scheduled maintenance. At approximately 1428 hrs. an automatic reactor trip occurred due to a Turbine Trip signal above P-9 (49% Reactor Thermal Power).
On a unit trip, station power is automatically transferred from the #2 Auxiliary Power Transformer (APT) to the 21 and 22 SPTs. Since the 22 SPT was unavailable, power was lost to the F and G group 4 KV non-vital busses. Undervoltage on the F and G busses caused a trip of the 23 and 24 Reactor Coolant Pumps (RCP).
The 21 and 22 motor driven, and the 23 turbine driven Auxiliary Feedwater (AFW) Pumps automatically started as expected on the unit trip due to low-low (14% Narrow Range) Steam Generator (SG) levels. At 1429 hrs. a second AFW actuation signal occurred due to low-low levels in the 21 and 22 SGs following the trip of the 23 and 24 RCPs. At 1457 hrs., SG low level setpoints were cleared and the 23 turbine driven AFW Pump was secured in accordance with operating procedures.
At 1523 hrs. a third AFW actuation signal occurred due to a low-low level in the 22 SG. The 21 and 22 motor driven AFW pumps remained in operation throughout unit stabilization.
The cause of the Turbine Trip signal was due to a spurious actuation of the Digital Electro-Hydraulic Controller (DEHC) {TG/HCU}. Investigation revealed that all three input channels of the DEHC were found to have simultaneously spiked above the overspeed trip setpoint of 103% of 1854 rpm to approximately 104% indicated. No actual change of turbine speed was observed.
Visual inspections were made of the DEHC and speed probes wiring, mounting and contacts to detect damage or looseness as well as evidence of foreign material intrusion. The turbine speed pickup tooth wheel was inspected for damage and alignment issues. Attempts were made locally at the DEHC and surrounding areas to reproduce false signals due to radio frequency interference. No issues were identified. A causal evaluation is in progress.
PREVIOUS OCCURRENCES
(272/2006-001) due to a spurious spiking of inputs to the DEHC causing a 103% overspeed turbine trip actuation.
SAFETY CONSEQUENCES AND IMPLICATIONS
Loss of forced reactor coolant flow is a Condition II Event analyzed in Section 15.2.5 of the UFSAR.
The UFSAR analysis states that the Departure from Nucleate Boiling (DNBR) will not decrease below the limiting value at any time during the transient and therefore no core safety limit is violated.
The UFSAR analysis assumes a reactor trip on low RCS loop flow with two loops coasting down.
Since reactor trip was actuated before the 23 and 24 loop low flow conditions occurred, the margin to Departure from Nucleate Boiling (DNB) was greater than that as analyzed in the UFSAR.
Therefore there were no safety consequences as a result of the loop low flow conditions due to the loss of the 23 and 24 RCPs.
Emergency operating procedures require operators to maintain SG levels above 9% Narrow Range level in at least one SG to ensure adequate inventory for secondary heat sink requirements. The third AFW actuation occurred with adequate inventory in the SGs, thus secondary heat sink requirements were never challenged during post-trip stabilization.
A review of this event determined that a Safety System Functional Failure (SSFF) as defined in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, did not occur.
CORRECTIVE ACTIONS
1. Salem is evaluating changes to the DEHC system to enhance its ability to withstand false spurious input signals.
2. The station has implemented a design change raising the existing Unit 2 DEHC system overspeed trip setpoint from 103% to 108% to reduce the likelihood of similar spurious actuations. A similar change is scheduled for implementation on Unit 1.
3. A causal evaluation is in progress to address additional corrective actions for the spurious turbine overspeed trip actuation.
COMMITMENTS
No commitments are made in this LER.