05000254/LER-2003-001
| Quad Cities Nuclear Power Station Unit 1 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown |
| 2542003001R00 - NRC Website | |
Quad Cities Nuclear Power Station Unit 1 05000254 ti (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
Reactor Shutdown due to Reactor Head Vent Steam Leak Constituting Pressure Boundary Leakage
A. CONDITION PRIOR TO EVENT
Unit: 1 � Event Date: May 20, 2003 � Event Time: 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> Reactor Mode: 2 � Mode Name: Startup � Power Level: 000% Startup (2) - Mode switch in Startup/Hot Standby position (or in Refuel position with all reactor vessel head closure bolts fully tensioned) with average reactor coolant temperature at any temperature.
B. DESCRIPTION OF EVENT
in the 2" Reactor Head Vent TSB] line. The leak was identified during the initial drywell entry for a reactor shutdown (Q1M16).
The Reactor Head Vent is utilized to vent off non-condensable gases from the reactor vessel head during operation. The vent line attaches to the vessel head at a 4" flanged connection and reduces to a 2" line constructed of A106 Grade B carbon steel, schedule 80, piping with socket welded fittings.
The reactor was subcritical in Mode 2 with a planned shutdown in progress and all Emergency Core Cooling (ECCS) systems operable. The individual performing the inspections discovered a small steam leak on the Unit 1 reactor head vent line. The leak was observed as a steam plume approximately 12 inches high. The leak was located in a section of piping inboard of the normally closed head vent isolation valves [ISV] and inboard of the normally open continuous head vent isolation valve.
The vent line and coupling [CPLG] are ASME Section XI Class 2 components. The leak was determined to be from a 60 degree circumferential crack in a fillet weld on a 2-inch coupling on the reactor head vent line. The weld is original construction (1970).
At 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br /> on May 20, 2003, a 4-hour Emergency Notification System notification was made per 10CFR50.72(b)(2)(i) for a reactor shutdown required by Technical Specifications due to reactor coolant pressure boundary leakage (Technical Specification 3.4.4, Condition C).
FACILITY NAME (1) PAGE (3) DOCKET NUMBER (2) LER NUMBER (6 Quad Cities Nuclear Power Station Unit 1 05000254 � (If more space Is required, use additional copies of NRC Form 366AX17)
C. CAUSE OF EVENT
The root cause of this event was inadequate verification of an original construction weld. The socket weld was poor quality as evidenced by significant porosity, lack of fusion and excessive overlap in the failure region. The defects propagated due to long-term corrosion and possibly fatigue to form the through-wall leak.
D. SAFETY ANALYSIS
The significance of the event was minimal. The leakage rate was low and all ECCS systems were operable.
The Drywell leakage (IJ] is monitored every four hours and a Drywell Continuous Air Monitoring system (IK) provides a Control Room alarm if particulate activity reaches a predetermined value. Also, the Drywell air temperature [IM] is trended daily in accordance with QCOS 1600-53.
The consequence of a postulated worse case failure is within analyzed limits. The area of the leaking line is less than 0.12 square feet even if it is doubled to account for the potential loss of reactor coolant from both ends of a postulated worst-case vent line break.
UFSAR Section 3.4.1.2.3, Protection of the Drywell and Torus, documents the evaluation of the internal flooding measures in the containment [NH], so that the worst-case failure of the reactor head vent line would not result in an accumulation of water beyond the analyzed limits.
UFSAR Section 15.6.5, Loss of Coolant Accidents Resulting from Piping Breaks Inside the Containment, documents the evaluation of the primary system piping failures, so that the consequences of the worst-case failure of the reactor head vent line would remain within the bounds of an analyzed small steam line break.
E. CORRECTIVE ACTIONS
Immediate Actions
The reactor shutdown that was in progress was continued to completion, and a 4-hour ENS notification was made.
An ultrasonic thickness examination of the line was performed to verify minimum wall thickness.
Corrective Actions Completed:
The failed weld, coupling and two-foot section of pipe were removed, and a new section of pipe and couplings was installed. The new welds were visually inspected and liquid penetrant tested.
line and on the similar coupling welds on Unit 2. No additional weld defects were identified.
DOCKET NUMBER (2) FACILRYNAMEO) (If more space Is required, use additional copies of NRC Form 366AX17)
F. PREVIOUS OCCURRENCES
LER NUMBER (6) PAGE (3) On February 27, 1998, during a Reactor Vessel Class. One Leak Test, a leak was identified in the heat affected zone of a coupling weld on the reactor vessel bottom head drain line. Failure analysis identified the failure as outside diameter initiated stress corrosion cracking. This event was reported in LER 1-98-012.
G. COMPONENT FAILURE DATA
There were no component failures associated with this event.