05000265/LER-2003-002
| Quad Cities Nuclear Power Station Unit 2 | |
| Event date: | |
|---|---|
| Report date: | |
| 2652003002R00 - NRC Website | |
DOCKET NUMBER (2) PAGE (3) FACILITY NAME (1) LER NUMBER (6 Quad Cities Nuclear Power Station Unit 2 05000265 2003 _ (If more space is required, use additional copies of NRC Form 366AX17)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
Self-Actuation of Main Steam Relief Valve due to Excessive Leakage Through Pilot Valve Seat
A. CONDITION PRIOR TO EVENT
Unit: 2 Event Date: April 16, 2003 Reactor Mode: 1 Mode Name: Power Operation Power Operation (1) - Mode switch in the RUN position temperature at any temperature.
Event Time: 1322 hours0.0153 days <br />0.367 hours <br />0.00219 weeks <br />5.03021e-4 months <br /> Power Level: 100% with average reactor coolant
B. DESCRIPTION OF EVENT
On April 16, 2003, at 1322 hours0.0153 days <br />0.367 hours <br />0.00219 weeks <br />5.03021e-4 months <br /> and with Unit 2 at 100% power, alarms associated with the Unit 2 3B Main Steam Relief Valve (RV) DM being open were received in the main control room. No activities involving the RVs were in progress at that time.
The operators initiated the appropriate procedures for an open RV. At 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br />, suppression pool cooling [BO] was initiated. At 1337 hours0.0155 days <br />0.371 hours <br />0.00221 weeks <br />5.087285e-4 months <br />, when the suppression pool temperature exceeded 95F, the operators manually scrammed the reactor. At 1340 hours0.0155 days <br />0.372 hours <br />0.00222 weeks <br />5.0987e-4 months <br /> the operators removed the control power fuses to the 3B RV, with no observed effect. At 1344 hours0.0156 days <br />0.373 hours <br />0.00222 weeks <br />5.11392e-4 months <br />, a Primary Containment isolation signal was received due to low reactor water level. At 1352, the operators closed the main steam isolation valves [ISV] to slow the reactor cooldown rate.
At 1359 hours0.0157 days <br />0.378 hours <br />0.00225 weeks <br />5.170995e-4 months <br />, the suppression pool temperature reached 110F and an Alert was declared in accordance with the Exelon Emergency Plan. At 1411 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.368855e-4 months <br />, the reactor cooldown rate exceeded the Technical Specification limit of 100F/hour.
At 2106 hours0.0244 days <br />0.585 hours <br />0.00348 weeks <br />8.01333e-4 months <br />, with reactor pressure at approximately 50 psig, the 3B RV indicated closed. At 2237 hours0.0259 days <br />0.621 hours <br />0.0037 weeks <br />8.511785e-4 months <br /> reactor coolant was below 212F and Mode 4, Cold Shutdown, was entered. At 2251 hours0.0261 days <br />0.625 hours <br />0.00372 weeks <br />8.565055e-4 months <br /> the Alert was exited.
C. CAUSE OF EVENT
The root cause of this event was excessive leakage past the 3B RV pilot valve seat, which allowed the RV closure force to diminish and the RV to fully open. The excessive leakage was due to erosion of the pilot valve seating surface as a result of an unknown pilot valve seat flaw. Possible causes of the seat flaw include on- line steam test actuations and very small initial leakage following refurbishment of the pilot valve seat.
DOCKET NUMBER (2) FACILITY NAME (1) PAGE (3) LER NUMBER (6) Quad Cities Nuclear Power Station Unit 2 05000265 � (If more space is required, use additional copies of NRC Form 366AX17) Contributing causes included a system monitoring plan that contained performance limits that failed to preclude inadvertent operation of the valve, and a RV design that is susceptible to a self-actuation due to a small amount of pilot valve leakage.
D. SAFETY ANALYSIS
The safety significance of this event was minimal. The spurious opening of the 3B RV did not affect the capability of the relief valves to protect against high reactor pressure. At the time of the event, the M Emergency Diesel Generator was out of service for planned maintenance. All other mitigating systems were available. Therefore, both emergency and non-emergency sources of injection to the vessel were available to make up for the coolant being relieved through the RV to the suppression pool.
The Probabilistic Risk Assessment of this event performed by Exelon risk personnel determined that the calculated conditional core damage probability was approximately 3E-7, which is below the level for very low risk (i.e., 1E-6).
E. CORRECTIVE ACTIONS
Corrective Actions Completed:
The failed Unit 2 3B RV was replaced.
The Unit 2 3B RV tailpipe thermocouple was relocated to the optimum location.
An analysis was performed to determine that the reactor coolant system was acceptable for continued operation following the cooldown rate of greater than 100F/hr.
Approval of a Technical Specification amendment to delete the requirement to actuate relief valves on-line has been obtained.
Relief from the ASME/OM Code requirement to lift relief valves at reduced pressure following maintenance has been obtained.
Corrective Actions to be Completed:
The relief valve post-maintenance (refurbishment) test requirements will be revised to include testing at a steam test facility at a nominal 1000 psi, with an acceptance criteria of a nominal zero lb/hr, in accordance with the vendor technical manual.
The System Engineering Main Steam monitoring plan will be verified or revised to include the appropriate tailpipe temperature monitoring frequency and action thresholds for relief valve tailpipe high temperatures.
FACILITY NAME (1) LER NUMBER (6) DOCKET NUMBER (2) PAGE (3) Quad Cities Nuclear Power Station Unit 2 05000265 (If more space is required, use additional copies of NRC Form 366AX17) QCOS 0203-02, "Safety And Relief Valve Temperature Surveillance," will be revised to ensure a Condition Report is initiated at a conservative tailpipe temperature to determine continued operation of a degraded relief valve.
F. PREVIOUS OCCURRENCES
No events were identified at Quad Cities Station involving self-actuation of a main steam relief valve.
G. COMPONENT FAILURE DATA
The Unit 2 3B main steam relief valve is a Model 93V-001 Power Operated Relief Valve manufactured by Target Rock.