05000336/LER-2006-002, Manual Reactor Trip Due to Trip of Both Feed Pumps Following a Loss of Instrument Air
| ML061220573 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/21/2006 |
| From: | Price J Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 06-284 LER 06-002-00 | |
| Download: ML061220573 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3362006002R00 - NRC Website | |
text
-i Dominion-Dominion Nuclear Connecticut, Inc.
Millstone Power Station Rope Ferry Road Waterford, CT 06385 APR 2 1 Mi U.',. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.
MPS Lic/WEB Docket No.
License No.06-284 RO 50-336 DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2 LICENSEE EVENT REPORT 2006-002-00 MANUAL REACTOR TRIP DUE TO TRIP FOLLOWING A LOSS OF INSTRUMENT AIR OF BOTH FEED PUMPSR This letter forwards Licensee Event Report (LER) 2006-002-00, documenting an event that occurred at Millstone Power Station Unit 2 on February 23, 2006. This LER is being submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual actuation of the Reactor Protection System (RPS).
If you have any questions or require additional information, please contact Mr.
David W. Dodson at (860) 447-1791, extension 2346.
Very truly yours, i
Al c Price ii(ce President - Millstone
--r ' 6;S
Serial No.06-284 LER 2006-002-00 Page 2 of 2 Attachments: (1)
Commitments made in this letter: None.
cc:
U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406-1415 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White flint North 11555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station Millstone Power Station Unit 2 LER 2006-002-00 Millstone Power Station Unit 2 Dominion Nuclear Connecticut, Inc. (DNC)
Abstract
On February 23, 2006, with the plant in Mode 1 and 100% power, a Manual Reactor Trip was initiated due to a trip of both feed pumps following a loss of Instrument Air. Mechanics were attempting to replace a pipe clamp on a two-inch copper Instrument Air header on Millstone Power Station Unit 2 when an inadequately soldered joint failed on a 1/2 inch tee connection. The 1/2 inch line separated from the header and resulted in rapidly lowering Instrument Air pressure in a portion of the turbine building and shutting of an excess flow check valve, as designed. Numerous air operated valves shifted to their loss-of-air failure position.
Feedwater heater high-level dump valves opened causing a reduction of heater drain flow and loss of suction pressure to the steam generator feed pumps (SGFP). Both SGFPs tripped and a manual reactor trip was initiated. The Auxiliary Feedwater System actuated as expected, however, the Number 1 Steam Generator AFW Regulating Valve went to its failed position (i.e. full open) allowing excessive flow to the Number 1 Steam Generator, which contributed to a greater than expected, cool down of the RCS. An operator was dispatched to take manual control of the regulating valve at which point RCS temperature was restored to the normal post-trip band of 530-5350 F. All other safety systems functioned as designed. The plant was stabilized in Mode 3 at normal operating temperature and pressure.
This event/condition is being reported pursuant to 50.73(a)(2)(iv)(A) as an event that resulted in the manual actuation of the Reactor Protection System as well as the automatic actuation of the Auxiliary Feedwater System.
NRC FORM 366 (6-2004)
(If more space Is required, use additional copies of NRC Forrn 366A) (17)
- 1. Event DescriDtion On February 23, 2006, with the plant in Mode 1 and 100% power, a Manual Reactor Trip was initiated following an Instrument Air (IA) [LD] leak that occurred while replacing a pipe clamp on a two-inch copper Instrument Air header in the Turbine Building. An inadequately soldered joint failed on a Y2 inch tee connection from a two-inch Instrument Air line that resulted in rapidly lowering Instrument Air pressure that caused the excess flow check valve to shut. Numerous air operated valves shifted to their loss-of-air position. Feedwater [SJ] heater high-level dump valves opened causing a reduction of heater drain flow and a loss of suction pressure to the Steam Generator [SB] Feed Pumps (SGFP). Both SGFPs tripped and a manual reactor trip was initiated. Non-Vital 120VAC Regulated AC panels VR1 1 and VR21 shifted to backup power supplies as expected due to the transfer of station power from the Normal Station Service Transformer (NSST) to the Reserve Station Service Transformer (RSST) following a reactor trip. The momentary loss of power to VR1 1 during this transfer resulted in a loss of letdown ard indication of Pressurizer Power Operated Relief Valve (PORV) and Main Steam Safety Valve (MSSV) position changes. Operators subsequently restored letdown and confirmed no actuation of either PORV's or MSSV's had occurred during the event.
Following the reactor trip, Control Element Assembly Position Display System (CEAPDS) indicated CEA 7 was not fully inserted and the Core Mimic indicated CEA 44 was not fully inserted. Upon further review, it was confirmed that both CEA 7 and 44 had fully inserted and the indication anomalies were due to reed switch indication behavior. Additionally following the trip, the plant experienced an abnormal cool down to 5260 F in part clue to excessive AFW flow to the Number 1 Steam Generator. An operator was dispatched to take manual control of the regulating valve at which point RCS temperature was restored to the normal post-trip band of 530-5350 F. It was subsequently determined that the Number 1 Steam Generator AFW Regulating Valve regulator was incorrectly set. This resulted in the Auxiliary Feed Regulating Valve going to its failed position (i.e., full open). All other safety systems functioned as designed.
The plant was stabilized in Mode 3 at normal operating temperature and pressure.
This event'condition is being reported pursuant to 50.73(a)(2)(iv)(A) as an event that resulted in the manual actuation cf the Reactor Protection System as well as the automatic actuation of the Auxiliary Feedwater System.
2. Cause
The cause of this event was determined to be an already weakened solder joint which was disturbed while attempting to repair an incorrectly installed clamp, not designed for Instrument Air piping. This in turn, resulted in a 1/2 inch copper line separating from a tee connection causing a partial loss of the Instrument Air System in the Turbine Building.
(If more space Is required, use additonal copies of NRC Form 366A) (17)
- 3. Assessment of Safety Conseauences There were no adverse consequences as a result of this event. The risk of the event was determined to be no greater than that of a manual reactor trip with a loss of the Main Feedwater. No loss of a safety function occurred.
The Auxiliary Feedwater remained available but was initially supplying excessive flow to the steam generators.
This was the result of the AFW Regulating Valve failing in the fully open position caused by an incorrect setting on the backup air regulator. After the setting was adjusted, the AFW flow control was restored and the RIDS cool down rate returned to normal. In summary, the trip is considered to be of low safety significance.
4. Corrective Actions
An investigation was conducted and appropriate corrective actions are being addressed in accordance with the Millstone Corrective Action Program.
The corrective actions to prevent recurrence of this condition are:
Training Review Board to determine necessary training for field workers; Instruct field workers to perform visual inspection of piping and to "snoop" (leak test) soldered joints within a close proximity prior to performing physical work on the Instrument Air system and; Inspect a sample of Instrument Air Piping and supports for proper components, installation and spacing.
5. Previous Occurrences
No previous similar events/conditions were identified.
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].