05000247/LER-2004-001, Re Manual Reactor Trip Due to Oscillating Feedwater Flow and Steam Generator Level with Flow Perturbations Caused by a Degraded Feed Water Regulating Valve
| ML043080311 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/25/2004 |
| From: | Dacimo F Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LER 04-001-00 | |
| Download: ML043080311 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv), System Actuation |
| 2472004001R00 - NRC Website | |
text
Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB O' n te RP.O.
Box 249 Buchanan, NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Vice President Administration October 25, 2004 Indian Point Unit No. 2 Docket Nos. 50-247 NL-04-127 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001
Subject:
Licensee Event Report # 2004-001-00, "Manual Reactor Trip Due to Oscillating Feedwater Flow and Steam Generator Level with Flow Perturbations Caused by a Degraded Feed Water Regulating Valve
Dear Sir:
The attached Licensee Event Report (LER) 2004-001-00 is the follow-up written report submitted in accordance with 10 CFR 50.73. This event is of the type defined in 10 CFR 50.73(a)(2)(iv) for an event recorded in the Entergy corrective action process as Condition Report CR-IP2-2004-04043.
There are no commitments contained in this letter. Should you or your staff have any questions regarding this matter, please contact Mr. Patric W. Conroy, Manager, Licensing, Indian Point Energy Center at (914) 734-6668.
Sincerely, Fred R. Dacimo Site Vice President Indian Point Energy Center
Docket Nos. 50-247 NL-04-127 Page 2 of 2 Attachment: LER-2004-001-00 cc:
Mr. Samuel J. Collins Regional Administrator - Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Resident Inspector's Office U.S. Nuclear Regulatory Commission Indian Point Unit 2 P.O. Box 59 Buchanan, NY 10511-0059 Mr. Paul Eddy State of New York Public Service Commission 3 Empire Plaza Albany, NY 12223-1350 INPO Record Center 700 Galleria Parkway Atlanta, Georga 30339-5957
Abstract
On September 1, 2004, at approximately 0005 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, Operations manually tripped the reactor as a result of oscillating Feedwater (FW) flow and 22 Steam Generator (SG) level with flow perturbations and FW pipe movement in the Auxiliary FW Pump Building.
All control rods fully inserted and all primary systems functioned properly.
The 22 FW flow control valve (FCV)-427 failed to fully close.
Operators initiated 22 FW isolation by closing the 22 FW isolation valves. At 0021 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, a 22 SG High-High level trip was actuated at 73t SG level initiating closure of the Main FW pump discharge valves and Main FW and Low Flow FW regulating and isolation valves. The plant was stabilized in hot standby with decay heat being removed by the main condenser.
Offsite power remained available and therefore the emergency diesel generators did not start.
The Auxiliary FW (AFW) system automatically started as a result of a SG low level normally experienced on trips from full power.
The cause of the event was a disengaged valve cage in FCV-427 from the valve body web.
The cause of the valve cage loosening was improper installation in 1997 due to inadequate guidance in the maintenance procedure used to verify that the cage was fully seated and properly torqued into the valve body web.
Significant corrective actions were inspection and repair of FCV-427 with revised guidance and revision of the valve maintenance procedure (AOV-B-012-A) to incorporate steps to verify that the cage is fully engaged and torqued into the valve body.
The event had no effect on public health and safety.
(If more space is required, use additional copies of NRC Fonn 366A) (17)
Note: The Energy Industry Identification System Codes are identified within brackets { }
DESCRIPTION OF EVENT
On September 1, 2004, at approximately 0005 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, while at 100W steady state reactor power, Operations manually tripped {JC} the reactor {RCT} as a result of oscillating Feedwater (FW) (SJj flow and 22 Steam Generator (SG) ISBI level with flow perturbations and FW pipe {P} movement in the Auxiliary FW (AFW) Pump Building {NF}.
Prior to the transient, on August 31, at 2348 hours0.0272 days <br />0.652 hours <br />0.00388 weeks <br />8.93414e-4 months <br />, while operating at 100% reactor power, with SG level control {JB} in AUTO, 22 SG narrow range (NR) level records show two cycles of level changes of approximately 2t and correction in automatic between 2348 hours0.0272 days <br />0.652 hours <br />0.00388 weeks <br />8.93414e-4 months <br /> and 2356 hours0.0273 days <br />0.654 hours <br />0.0039 weeks <br />8.96458e-4 months <br /> with no operator action.
Subsequently, at 2356 hours0.0273 days <br />0.654 hours <br />0.0039 weeks <br />8.96458e-4 months <br />, operators observed 22 SG NR level starting to decrease from a normal value of 49% to 30% with a St deviation alarm annunciated at 44%.
CR operators observed oscillating FW flow and erratic behavior of the 22 Main FW regulating valve FCV-427 {FCVI. At 0001 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, Operators entered Abnormal Operating Procedure 2AOP-FW-1 and placed the FW regulating valve (FCV-427) in manual and attempted to increase FW flow in 22 SG without success.
Excessive FW flow oscillations continued.
Operators then opened low flow bypass valve FCV-427L to increase SG level which started 22 SG level increasing at a level of 30t. At approximately 35% SG level valve FCV-427L was returned to closed. At approximately 0004 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, a Nuclear Plant Operator (NPO) in the AFW Pump Building reported to the control room loud noises due to flow perturbations and pipe movement.
Based on plant conditions, the Control Room Supervisor (CRS) directed a manual reactor trip (RT) {JC}.
All control rods {AA} fully inserted and all primary systems functioned properly.
The 22 FW regulating valve FCV-427 failed to fully close.
Operators initiated FW isolation by closing FW motor operated isolation valves (MOV) BFD-5-1 {ISV}
and BFD-90-1 {ISV}.
At 0021 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, a 22 SG high level trip {JB} was actuated at 73% SG level, initiating automatic closure of the Main FW Pump motor operated discharge valves (BFD-2-21 and BFD-2-22), Main FW and Low Flow FW regulating and isolation valves, and trip of the turbine driven Main FW Pumps.
The plant was stabilized in hot standby with decay heat being removed by the main condenser
{SG}.
Offsite power remained available and therefore the emergency diesel generators {EK} did not start.
The AFW System {BA} automatically started as a result of a SG low level normally experienced on trips from full power.
FW regulating valve FCV-427 is a Copes-Vulcan {C635} globe valve {V} with Copes-Vulcan actuator Model D-1000-160.
The valve has a positioner to perform its modulating function and 3 solenoids {SOL} attached to the actuator for fast closure.
CR operators observed the rod bottom lights, Reactor Trip (RT) First Out Annunciator (Manual Trip).
The plant was stabilized in hot standby with decay heat being released to the main condenser via the steam dump valves {V}.
(If more space is required, use additonal 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)
After reactor shutdown, the 22 SG reached the SG High-High level set point (73%)
and in accordance with plant design the proper actuation signals were initiated to isolate FW addition.
Plant design is for FW/SGL to automatically isolate to preclude excessive RCS cooldown, containment overpressure, and SG overfill.
FW/SGL isolation is initiated by a Hi-Hi SGL signal or a safety injection signal.
A SG high-high level signal to the FW/SGL control system on two-out-of-three high SGL in any one of four SGs initiates FW isolation.
The protection signals provide redundant isolation.
Redundant FW isolation is accomplished by automatically closing all main and bypass FW control valves and closing the FW Pump discharge isolation valves.
The closure of the FW Pump discharge isolation valves will automatically trip the FW Pumps and close the motor-operated isolation valves upstream of the FW control valves.
The SG Hi-Hi level trip also initiates Main Generator trip (86P and 86BU relays)/TT.
For this event the manual RT initiated a TT/Main Generator trip therefore the RT/TT actuation had already been completed when the SGL Hi-Hi level actuation occurred.
This event was bounded by the analyzed event described in FSAR
- Section 14.1.10, Excessive heat removal due to a FW system malfunction. The plant performed as expected and the event was bounded by the FSAR analysis. For this event rod control was in automatic and the reactor scrammed immediately upon a manual reactor trip.
RCS pressure remained below the set point for pressurizer PORV or code safety valve operation and above the set point for automatic safety injection actuation.
Following the RT, the plant was stabilized in hot standby.