05000334/LER-2003-001
| Beaver Valley Power Station Unit No. 1 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation |
| 3342003001R00 - NRC Website | |
FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)
PLANT AND SYSTEM IDENTIFICATION
Westinghouse-Pressurized Water Reactor (PWR) Reactor Protection System (JC) Engineered Safety Features Actuation System Instrumentation (JE) Main Steam System (SB) High Pressure Safety Injection System (BQ) Auxiliary Feedwater System (BA) Emergency Onsite Power Supply (EK)
CONDITIONS PRIOR TO OCCURRENCE
Unit 1: Mode 1 at 100 % power There were no systems, structures, or components that were inoperable that contributed to the event other than as described below.
DESCRIPTION OF EVENT
On 2/24/03 at 1548 hours0.0179 days <br />0.43 hours <br />0.00256 weeks <br />5.89014e-4 months <br />, Beaver Valley Power Station (BVPS) Unit 1 was operating at 100% power, end of core life at steady state conditions, with preliminary Refueling No. 15 work evolutions (scaffold erection) in progress, when the Control Room received Annunciator A1-56 "Steamline Stop Valve Not Fully Open". This alarm was followed by a safety injection and reactor trip due to low main steam pressure.
The alarms were the result of the 'C' Main Steam Isolation Valve closing, which caused a Low Steam Line Pressure safety injection and reactor trip. The remaining two main steam isolation valves automatically closed due to the steam pressure decrease rate. As a result of the safety injection, the high head safety injection pumps were automatically realigned to inject into the Reactor Coolant System (RCS) and both emergency diesel generators started but did not load since offsite power remained available. As a result of the safety injection and reactor trip, the auxiliary feedwater pumps automatically started and commenced providing feedwater to the steam generators. Control room personnel implemented Emergency Operating Procedure E-0, "Reactor Trip / Safety Injection". All required safety related systems functioned as required. After the automatic actuation occurred, control room operators noted that all three steam generator pressures were stable at normal post reactor trip values. No indications of a steam leak were observed. The plant was stabilized in Mode 3.
Per the Emergency Plan, at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, an Unusual Event was declared. At 1605 hours0.0186 days <br />0.446 hours <br />0.00265 weeks <br />6.107025e-4 months <br />, the safety injection was terminated per procedure, E.S.1.1, Safety Injection Termination. At 1635 hours0.0189 days <br />0.454 hours <br />0.0027 weeks <br />6.221175e-4 months <br />, procedure FR-I.1, Response To High Pressurizer Level was performed to re-establish a pressurizer steam bubble. At 1658 hours0.0192 days <br />0.461 hours <br />0.00274 weeks <br />6.30869e-4 months <br />, operations exited from procedure FR-I.1, Response To High Pressurizer Level, and completed the remaining steps of E.S.1.1. At 1701 hours0.0197 days <br />0.473 hours <br />0.00281 weeks <br />6.472305e-4 months <br />, procedure FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3) DESCRIPTION OF EVENT (Continued) 10M-11.4M, Recovery Procedure From Safety Injection procedure was performed. At 1735 hours0.0201 days <br />0.482 hours <br />0.00287 weeks <br />6.601675e-4 months <br />, the Unusual Event was terminated.
REPORTABILITY
An automatic safety injection and reactor trip occurred on February 24, 2003. This resulted in an automatic actuation of the Reactor Protection System (reactor trip), the main steam isolation system (multiple main steam isolation valves), the emergency core cooling system (ECCS) (high head safety injection pumps), the auxiliary feedwater system (motor driven and steam driven auxiliary feedwater pumps), and the emergency ac electrical power system (onsite emergency diesel generators). This event is reportable pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of systems listed in paragraph (a)(2)(iv)(B)(1), (a)(2)(iv)(B)(2), (a)(2)(iv)(B)(3), and (a)(2)(iv)(B)(6), and (a)(2)(iv)(B)(8). The NRC was notified that an automatic safety injection and reactor trip occurred at BVPS Unit 1 which led to an Unusual Event declaration which was reported pursuant to 10 CFR 50.72(a)(1) at 1656 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.30108e-4 months <br /> on February 24, 2003 (ENS 39616). In addition, this event was reported pursuant to 10 CFR 50.72 (b)(2)(iv)(A), (b)(2)(iv)(B) and (b)(3)(iv)(A). This LER also satisfies Technical Specification 3.5.2.b which requires a special report to be submitted to the NRC describing the circumstances of an ECCS actuation that injects water into the RCS and the total accumulated actuation cycles to date.
CAUSE OF EVENT
The cause for the safety injection and reactor trip was the sudden inadvertent closure of the 'C' Main Steam Isolation Valve (MSIV). The west air cylinder rupture disk on the MSIV was damaged due to human error by craft workers installing scaffolding in the area below the MSIV. While maneuvering a scaffold pole into position, a craft worker positioned it through a gap in the overhead grating. The top of the pole then punctured a rupture disk on one of the two air operating cylinders that hold the MSIV in the open position. The rupture disk failed, thereby venting the instrument air pressure resulting in closure of the MSIV.
ANALYSIS OF EVENT
The MSIVs are configured like check valves installed in the reverse direction in the main steam lines.
The valves are held open by two pneumatic cylinders, one on each side of the valve. Puncturing the rupture disk and the subsequent venting of its instrument air caused the 'C' MSIV to close, isolating steam flow in the C loop. Steam flows in the A and B loops subsequently increased to meet the steam demand of the turbine, which consequently resulted in a steam pressure drop in the B loop such that reactor trip, safety injection and steam line isolation signals were generated. Although the actual B loop steam line pressure never decreased to the low steam line pressure safety injection setpoint of 500 psig, the lead/lag sensitivity in the ESF actuation instrumentation caused an automatic safety injection signal to be generated as a result of the rate of the pressure decrease.
ANALYSIS OF EVENT (Continued) The post-trip review did identify a MSIV configuration control error since no debris-catcher basket was attached to outside of the rupture disc as shown on design drawings. However, the debris basket is not necessary for valve performance.
The total accumulated actuation cycles to date at BVPS Unit No. 1 which involved injecting water into the Reactor Coolant System following a Safety Injection signal in Modes 1-4 is 23. This includes two events during pre-operational testing. There are an additional 22 inadvertent safety injection actuation cycles where no water was injected into the Reactor Coolant System or the safety injection occurred in Mode 5-6.
SAFETY IMPLICATIONS
The plant risk associated with the BVPS Unit 1 safety injection and reactor trip when the main steam isolation valve 'C' MSIV closed on 02/24/2003, is considered to be low. This is based on the conditional core damage probability for the event when considering the actual component unavailabilities that were present at the time of the trip. Therefore, the safety significance of this event was small.
CORRECTIVE ACTIONS
1. The rupture disks on each of the MSIVs were replaced.
2. Short term administrative controls were established to obtain Shift Manager permission prior to any work in the Main Steam Valve Room during plant operation.
3. A design change has been initiated to install cover plates on the grating at the MSIVs to provide physical protection of the rupture disks for the MSIV actuators.
4. The standard for the scaffold erection process was revised to have SRO-qualified Operations personnel perform walkdowns prior to erecting scaffolding to identify potential concerns and precautions associated with on-line equipment.
5. Self checking techniques and expectations were reinforced with site craft personnel. Specific emphasis was placed on craft responsibility to prevent inadvertent contact with plant operating equipment.
Completion of the above and other corrective actions are being tracked through the corrective action program.
FACILITY NAME (1) LER NUMBER (6) DOCKET (2) PAGE (3)
PREVIOUS SIMILAR EVENTS
A review of past Beaver Valley Power Station Licensee Event Reports for the last five years found two similar events involving an inadvertent/unintentional Engineered Safety Feature (ESF) actuation at BVPS Unit 1 or Unit 2. There have been no other safety injection actuations in the last five years.
BVPS Unit 1 LER 00-004, "Inadvertent ESF Actuation Due to Loss of Power to 4kv Emergency Bus.
BVPS Unit 2 LER 00-001, "ESF Actuation of Feedwater Isolation While Shutting the Plant Down for Refueling.