05000352/LER-2011-002
| Docket Number | |
| Event date: | 0-3-2011 |
|---|---|
| Report date: | 08-02-2011 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3522011002R00 - NRC Website | |
CONTINUATION SHEET
Unit Conditions Prior to the Event Unit 1 was in Operational approximately 100% power.
components out of service
Description of the Event
Condition (OPCON) 1 (Power Operation) at There were no structures, systems or that contributed to this event.
On Friday, June 3, 2011, Limerick Unit 1 was operating at 100% power. The quarterly "Feedwater/Main Turbine Trip System Actuation - Reactor Vessel Water Level - High Level 8; Channel B Functional Test" (ST-2-042-634-1) surveillance was in-progress. The step that verifies that the opposite channel trip contacts were open was being performed by measuring the voltage difference across each contact with a volt-ohm meter (VOM). At 1021 hours0.0118 days <br />0.284 hours <br />0.00169 weeks <br />3.884905e-4 months <br />, a main turbine trip (EIIS:JJ) occurred due to an invalid reactor vessel high level trip actuation. The turbine trip caused an automatic actuation of the reactor protection system (RPS) (EIIS:JC).
The operators entered the transient response and operating procedure for reactor pressure vessel (RPV) control (T-101) and stabilized reactor parameters. The operators verified that all control rods were fully inserted.
Reactor level initially decreased to a minimum of zero inches and then increased to a maximum of plus 70 inches. The reactor feed pumps (EIIS:SJ) tripped when level exceeded the plus 54 inches setpoint. The reactor level of less than +12.5 inches resulted in an isolation signal to the closed Group IIB valves as expected.
Reactor pressure initially increased to approximately 1163 psig.
Reactor pressure remained less than the lowest safety relief valve (SRV) setpoint of 1170 psig; therefore, no SRVs actuated. The main steam bypass valves opened as designed to control pressure. All turbine supervisory functions and generator protective relaying functioned as designed.
The post-scram troubleshooting identified that the high level trip surveillance test caused the invalid high level trip actuation. The test verification that the other channel trip logic contacts were open by reading DC voltage at the contacts using a VOM was in progress when the actuation occurred.
A 4-hour NRC ENS notification was required by 10CFR50.72(b)(2)(iv)(B) for an actuation of RPS when the reactor was critical. An 8-hour NRC ENS notification was required by 10CFR50.72(b)(3)(iv)(A) for a valid actuation of RPS. The ENS notification (#46919) was completed on Friday, June 3, 2011, at 12:06 ET. This event involved an automatic actuation of RPS.
Therefore, this LER is being submitted pursuant to the requirements of 10CFR50.73(a)(2)(iv)(A).
Analysis of the Event
There was no actual safety consequence associated with this event.
The potential safety consequences of this event were minimal. The operators effectively stabilized reactor parameters after the reactor vessel high level trip.
A Simpson 260 VOM was used to verify that all four high level trip logic channel contacts were open prior to testing the channel nEln trip logic. The VOM is used to measure DC voltage at each contact to verify a nominal zero voltage difference. The investigation determined that as the VOM was ranged down the logic tripped when the 2.5 VDC range was selected.
The tests were revised in 2004 and the test equipment was changed at that time. The technical error was a historical issue and created a latent risk.
Cause of the Event
The root cause of the event was the reactor vessel high level trip calibration and functional surveillance test revisions did not fully assess the impacts of the test equipment on the DC turbine trip circuit.
Corrective Action Planned The procedure for the control of portable measurement and test equipment program (MA-AA-716-040) will be revised to include a requirement to evaluate changes to the type of test equipment used and test methodology.
The reactor vessel high level trip calibration and functional surveillance tests will be revised to incorporate test equipment that is compatible with the DFWLCS prior to the next performance.
Previous Similar Occurrences There were no previous similar occurrences in the prior three years.
Component data:
System:
Component:
Manufacturer:
Model:
Electro-hydraulic Control 10-C663 Panel, EHC, Turbine G080 General Electric Company 838E408