05000244/LER-2003-002, Major Power Grid Disturbance Causes Loss of Electrical Load and Reactor Trip
| ML032890441 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/09/2003 |
| From: | Mecredy R Rochester Gas & Electric Corp |
| To: | Clark R Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 03-002-00 | |
| Download: ML032890441 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| 2442003002R00 - NRC Website | |
text
Robert &. Mecredy REnff Vice President Always at Your Service Nuclear Operations October 9, 2003 Mr. Robert L. Clark Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555
Subject:
LER 2003-002, Major Power Grid Disturbance Causes Loss of Electrical Load and Reactor Trip R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Clark:
The attached Licensee Event Report (LER) 2003-002 is submitted in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv)(A).
This event has in no way affected the public's health and safety.
Ver~y~y yours, C. Mecedy /
y xc:
Mr. Robert L. Clark (Mail Stop 0-8-C2)
Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector An equal opportunity employer L
89 East Avenue I Rochester, NY 14649 tel (585) 546-2700 lo)
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www.rge.com An Energy East Company
Abstract
On August 14, 2003, the plant was in Mode 1 at approximately 100% steady state reactor power. At approximately 1611 EDST, a major electrical grid disturbance occurred, affecting the Northeastern United States and areas of Southern Canada. The disturbance caused a complete loss of electrical load and an automatic Reactor Trip. The Shift Supervisor conservatively declared an Unusual Event at 1646 EDST based on Emergency Action Level 6.1.1, "Loss of ability to supply power to the safeguard trains from offsite circuits 751 and 767 for greater than 15 minutes," and exited the Unusual Event at 2108 EDST.
The cause of the Reactor Trip was 2/4 Over Temperature Delta T (OTDT) channels reaching their setpoints due to the load rejection and ensuing Reactor Coolant System transient.
NRCm FRM 300 if-Am )
(ff more space is required, use additional copies of NRC Fonn 366A)
I.
PRE-EVENT PLANT CONDITIONS:
On August 14, 2003 the plant was in Mode 1 at approximately 100% steady state reactor power.
Turbine Generator Volt-Amps Reactive (VAR) testing had been performed earlier in the day and the systems had been returned to normal operation. There were no electrical grid problems apparent from the Control Room and the operators had not been notified of any problems by RG&E Energy Operations.
II.
DESCRIPTION OF EVENT
A.
EVENT:
At approximately 1611 EDST on August 14, 2003 the Control Room received several alarms associated with a loss of main generator electrical load and the associated plant transient. The plant initially experienced a turbine runback as the control systems responded to the loss of electrical load. Both Pressurizer Power Operated Relief Valves (PORVs) lifted and re-closed per design to limit the reactor Coolant System pressure transient.
Approximately 28 seconds after receiving the first indication of a transient, the Reactor automatically tripped on Over Temperature Delta T (OTDT). The main generator tripped as designed due to the reactor/turbine trips.
Subsequent to the trip, Main Feedwater Isolation occurred as designed on low Tavg coincident with a reactor trip. However, due to voltage swings from the grid disturbance, instrument variations caused the Advanced Digital Feedwater Control System (ADFCS) to transfer to manual control. This transfer overrode the isolation signal causing the Main Feedwater Regulating Valves (MFRVs) to go to, and remain at, the normal or nominal automatic valve demand position at the time of the transfer, resulting in an unnecessary feedwater addition. The feedwater addition was terminated when the MFRVs closed on the high-high steam generator level (85%) signal. Although indicated level continued to increase due to overshoot and heatup of the water in the steam generators, subsequent evaluation of the level trends and walk down of the main steam header supports by Engineering personnel determined that an overfill condition did not occur.
The Main Steam Isolation Valves (MSIVs) were manually closed per procedure FR-H.3, "Response to Steam Generator High Level." This event does not meet the definition for NRC Performance Indicator (PI) "scram with loss of normal heat removal." A Frequently Asked Question (FAQ) has been submitted to clarify this position.
At approximately 1635, both Reactor Coolant Pumps (RCPs) tripped on Under Frequency (UF) as designed due to continued grid disturbances. The plant was stabilized in Mode 3 in natural circulation. Subsequent to the trip of the RCPs, the #2 seal on both pumps opened unc.- 9:ngRA 1190th f4_1"4%
(if more space Is required, use additional copies of NRC Form 3664) causing #1 seal return flow to decrease to zero. After consultation with Westinghouse, the RCPs were successfully restarted at 0437 and 0530 EDST on August 15, 2003. The seal flow indications then returned to normal values. The plant is designed to operate in natural circulation in Mode 3 with heat removal through the steam generators.
While off-site power was lost to several on site buildings, off-site power was never lost to the busses supplying the power block area. Although the safeguards bus voltage was swinging as a result of the grid transient, the voltage did not reach the undervoltage setpoints. However, the Operators determined that the off-site supply was unreliable, manually started the Emergency Diesel Generators (EDGs), and then manually transferred the safeguards busses to the EDGs. This resulted in declaring an Unusual Event that was terminated after power was later transferred back to the off-site sources.
When the Main Feedwater Pumps were stopped per procedure, the Motor Driven Aux Feedwater (MDAFW) pumps started coincident with the existing turbine trip signal. During subsequent recovery operations the B MDAFW Pump was damaged due to an error in pump alignment. The error was the result of a missed procedure step.
Seventy two (72) of ninety six (96) sirens associated with the Ginna prompt notification system were out of service due to power outages across the RG&E system, and were returned to service as power was restored to the respective service areas by August 15, 2003.
B.
INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
None C.
DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES
August 14, 2003, 1611 EDST: Event Date and Time, Onset of grid disturbance, PORVs cycle, and Reactor Trip.
August 14, 2003, 1612 EDST: Main Generator Trip August 14, 2003, 1631 EDST: Main Steam Isolation Valves Closed August 14, 2003, 1635 EDST: Both RCPs trip, plant enters natural circulation.
August 14, 2003, 1646 EDST: Declaration/Notification of Unusual Event, event
- 40068, Under 10CFR 50.72a(l)(i)
(ff more space Is required, use additional copies of (if more space Is required, use additional copies of NRC Fonn 366A)
(1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or (2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed."
As an assessment considering the consequences and implications of this event, a review of the UFSAR transients was conducted. For evaluation of analyzed transients versus this event, the UFSAR specifically analyzes a loss of external electrical load in Section 15.2.2. Four cases are considered, two at beginning of life (BOL) and two at end of life (EOL), with and without automatic pressurizer pressure control. The case analyzed for EOL with automatic pressure control would normally be expected to more closely reflect the actual transient. However, the BOL case with automatic pressure control was more representative. This is due to the fact that the UFSAR assumes no rod control and loss of main feedwater (MFW) coincident with the trip. This maximizes the reactivity feedback while resulting in a reactor trip on low steam generator level.
Since both automatic rod control and MFW functioned, reactivity feedback was minimized and the reactor tripped on OTDT similar to the BOL transient analyzed. No items of concern were noted.
Both Emergency Diesel Generators operated as designed throughout the event, ensuring a reliable source of power to the AC emergency busses at all times.
The re-initiation of MFW due to the ADFCS transfer to manual was evaluated for safety concerns.
The FW isolation signal from low Tavg coincident with a reactor trip is a control signal rather than a required safety function. Although the unexpected bypass of the signal caused a control problem, the safety related high high level feedwater isolation signal functioned as required, preventing a SG overfill condition. Since the safety related high-high steam generator level feedwater isolation functioned as required, and the steam generators did not enter an over fill condition, there was no compromise of safety as a result of this event.
Regarding the damaged B MDAFW pump, Ginna has a total of four 100% capacity MDAFW pumps, including two Standby Auxiliary Feedwater (SAFW) pumps and one 200% capacity Turbine Driven Auxiliary Feedwater (TDAFW) pump.
Therefore it was determined that the plant responded within it's design and licensing basis, that there were no unreviewed safety questions, and that the public's health and safety was assured at all times.
(If more space Is required, use additional copies of (If more space Is required, use additional copies of NRC Form 366A)
VI.
ADDITIONAL INFORMATION
A.
FAILED COMPONENTS:
B MDAFW pump - Worthington Model # 2 WTF-87 B.
PREVIOUS LERs ON SIMILAR EVENTS:
An historical search of LERs was conducted with the following results: There are no similar LERs where power grid disturbances resulted in a reactor trip.
C.
THE ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) COMPONENT FUNCTION IDENTIFIER AND SYSTEM NAME OF EACH COMPONENT OR SYSTEM REFERRED TO IN THIS LER:
COMPONENT IEEE 803 IEEE 805 FUNCTION IDENTIFIER SYSTEM IDENTIFICATION Auxiliary Feedwater Pumps P
BA Main Feedwater Regulating Valves FCV Si Main Steam Isolation Valves ISV SB Reactor Coolant Pump P
AB Emergency Diesel Generators DG EK D.
SPECIAL COMMENTS:
None WR('. 1:QPM qrAA II-9MI%