05000440/LER-2003-002, Regarding Reactor Scram as a Result of a Loss of Off-Site Power
| ML032950401 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 10/14/2003 |
| From: | Kanda W FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PY-CEI/NRR-2737L LER 03-002-00 | |
| Download: ML032950401 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| 4402003002R00 - NRC Website | |
text
FENOC FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant 10 Center Road Perry, Ohio 44081 W1iharn R. Kanda Vice President - Nuclear 440-280-5579 Fax: 440-280-8029 October 14, 2003 PY-CEI/NRR-2737L United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Perry Nuclear Power Plant Docket No. 50-440 LER 2003-002-00 Ladies and Gentlemen:
Enclosed is Licensee Event Report (LER) 2003-002, Reactor Scram as a Result of a Loss Of Off-Site Power. This event is being reported in accordance with 10CFR50.73.
There are no regulatory commitments contained in this letter. Any actions discussed in this document that represent intended or planned actions, are described for the NRC's information, and are not regulatory commitments.
If you have questions or require additional information, please contact Mr. Vernon K.
Higaki, Manager - Regulatory Affairs, at (440) 280-5294.
- - Very trulyyours --
7 Enclosure: LER 2003-002 cc: NRC Project Manager NRC Resident Inspector NRC Region IlIl
,5
Abstract
On August 14, 2003, at 1610 hours0.0186 days <br />0.447 hours <br />0.00266 weeks <br />6.12605e-4 months <br />, with the Perry Nuclear Power Plant operating in Mode 1, at approximately 100 percent reactor power, a generator trip due to under-frequency occurred resulting in a turbine control valve fast closure scram. The under-frequency was the result of a loss of off-site power. The cause of the grid disturbance that resulted in Ahe16s-sof-(ff-sit§eiowesr is -still under idv-e-1iW ti-by-F rstErergy.-The loss f-off-site power req0ired-ehtr7iintd-the emergency plan for an Unusual Event. Entry into the emergency plan is reportable per 10CFR50.72(a)(1)(i). Additional reporting requirements are documented in the event report.
Loss of off-site power to the safety-related emergency power supply busses for greater than 15 minutes resulted in the declaration of an Unusual Event, which was reported to the NRC via the Emergency Notification System (ENS) at 1635 hours0.0189 days <br />0.454 hours <br />0.0027 weeks <br />6.221175e-4 months <br />. Subsequently, it was recognized that power was not available at the Emergency Operations Facility and the Backup Emergency Operations Facility. This condition was reported to the NRC via the ENS at 2225 hours0.0258 days <br />0.618 hours <br />0.00368 weeks <br />8.466125e-4 months <br />. The Unusual Event was terminated at 1952 hours0.0226 days <br />0.542 hours <br />0.00323 weeks <br />7.42736e-4 months <br /> on August 15, 2003, following restoration of off-site power. Following restoration of power to the emergency busses with the emergency diesel generators, low pressure alarms were received on low pressure core spray (LPCS) and on residual heat removal (RHR) A loop. Following initial investigation, both LPCS and RHR A were declared inoperable at 1847 hours0.0214 days <br />0.513 hours <br />0.00305 weeks <br />7.027835e-4 months <br />. The associated water-leg pump was determined to be air bound and was vented. Both LPCS and RHR A were vented prior to returning to service.
This report also satisfies Operational Requirements Manual section 7.6.2.1, which requires a special report submittal following an Emergency Core Cooling System actuation and injection into the reactor coolant system.
NRC FORM 366 (7-2001)
(If more space is required, use additional copies of (if more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of (If more space is required, use additional copies of NRC Form 366A) (17) radiation monitor. The radiation monitor provides isolation signals to the off-gas system on both hi radiation and a downscale signal. This condition had no impact since the main steam isolation valves received an isolation signal and closed about 4 seconds later.
IV. SAFETY ANALYSIS
This event, a LOOP, was compared to the loss of all grid connection event that is evaluated in the updated safety analysis report (USAR), section 15.2.6. Note that a LOOP, although a safety significant event, is less severe than the station blackout (SBO) event that is evaluated in USAR Appendix 15H. The loss of all grid connection event is categorized as an incident of moderate frequency, potential to occur about every 20 years. The sequence of actual events followed.those-asilisted inlShe-USAR.-An exception was-that-SRVs opened-later-as a result of aslower-- --
pressure increase (about 9 seconds after MSIV isolation). The conditional core damage probability (CCDP) calculated for the loss of all grid event was 1.25E-04.
The LPCS/RHR A water-leg pump air binding event of August 14, 2003, resulted in the RHR A and the LPCS pumps being declared inoperable. Had a design bases event LOOP, concurrent with a loss of coolant accident (LOCA) occurred, the air binding would have still been localized to the water-leg pump piping and would have had no impact on the LPCS or RHR A pumps. The LPCS and RHR A pumps would have automatically started upon receipt of the LOCA initiation signal prior to injection system depressurization. (The impact is currently being confirmed in PNPPs corrective action program and will be communicated in a supplemental report.) Additionally, the LPCS and RHR A pumps are just 2 of several sources of makeup that can be used during a LOOP event. This condition had no significant impact on the event (LOOP). Due to the difference in piping configuration, the division 2 (RHR B and C), HPCS and RCIC water-leg pumps are not affected.
V, CORRECTIVE ACTIONS The grid instability that caused the generator trip and resulted in the reactor scram is still being reviewed by FirstEnergy. Lessons learned by the industry that apply to PNPP will be reviewed for incorporation. The operation of the turbine, generator and main transformer were assessed. No damage to this equipment was identified to have occurred as a result of this event.
The LPCS/RHR A water-leg pump and the feedwater leakage control piping was vented and returned to normal operation. Initial interim measures were to vent the piping weekly until the appropriate interval could be determined.
The piping is currently being vented at two-week intervals. Procedures were revised to add appropriate vent points.
Design initiatives will be explored to modify the FWLCS piping to'avoid or reduce the-potential for-air-collection_-.-------
within the division I FWLCS piping.
The combustible gas drywell purge inboard containment isolation valve relays, K36 and K39, were cycled and the circuitry functioned as designed. The condition was entered in the corrective action program and the relays were conservatively scheduled for replacement.
Additional backup facilities were identified during the event that could have been used as an EOF during the event.
The inability to obtain a timely reactor water sample was entered into the corrective action process for trending purposes. No additional actions are required.
VI. PREVIOUS SIMILAR EVENTS
No previous loss of off-site power event or pump air binding event was identified to have occurred at the PNPP.
Energy Industry Identification System (EIIS) codes are identified in the text in the format [xx].