05000254/LER-2003-001, Reactor Shutdown Due to Reactor Head Vent Steam Leak Consulting Pressure Boundary Leakage
| ML032120510 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/21/2003 |
| From: | Tulon T Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SVP-03-084 LER 03-001-00 | |
| Download: ML032120510 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown |
| 2542003001R00 - NRC Website | |
text
Exe lk, n W Exelon Generation Company, LLC www.exeloncorp.com Nuclear Quad Cities Nuclear Power Station 22710 206l Avenue North Cordova, IL 61242-9740 July 21, 2003 SVP-03-084 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Quad Cities Nuclear Power Station, Unit 1 Facility Operating License No. DPR-29 NRC Docket No. 50-254
Subject:
Licensee Event Report 254/03-001, "Reactor Shutdown due to Reactor Head Vent Steam Leak Constituting Pressure Boundary Leakage."
Enclosed is Licensee Event Report (LER) 254/03-001, "Reactor Shutdown due to Reactor Head Vent Steam Leak Constituting Pressure Boundary Leakage," for Quad Cities Nuclear Power Station.
This report is submitted in accordance with the requirements of the Code of Federal Regulations, Title 10, Part 50.73(a)(2)(i)(A), which requires reporting of the completion of any nuclear plant shutdown required by the plant's Technical Specifications.
Should you have any questions concerning this report, please contact Mr. W. J. Beck at (309) 227-2800.
Respectfully, 6tothy J. Tulon Site Vice President Quad Cities Nuclear Power Station cc:
Regional Administrator - NRC Region IlIl NRC Senior Resident Inspector - Quad Cities Nuclear Power Station
I
Abstract
On May 20, 2003, at approximately 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />, a small (approximately 12" plume) steam leak was identified on the Unit 1 Reactor Head Vent line.
At the time that the leak was identified, the reactor was subcritical in Mode 2 with a planned shutdown in progress and all Emergency Core Cooling ECCS) systems operable. The leak was identified during the initial drywell entry for the reactor shutdown.
The Reactor Head Vent is utilized to vent 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 leak was located adjacent to an original construction weld in the 2 section of piping inboard of the isolation valves.
The root cause of this event was inadequate verification of weld quality. The significance of the event was minimal.
The leakage rate was low and all ECCS systems were operable.
Also, the consequences of a postulated worse case failure of this pipe are within analyzed limits. Corrective actions included replacement of a section of piping and inspection of additional welds on Unit 1 and Unit 2.
NRC PORM 366A (7-2001)U.S. NUCLEAR REGULATORY COMMISSION (7201)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIALTIREVISION Quad Cities Nuclear Power Station Unit 1 05000254 NUMBER NUMBER 2003 001 00 2 of 4 (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
On May 20, 2003, at approximately 0240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />, a steam leak was identified on Unit 1 in the 2 Reactor Head Vent SB] line.
The leak was identified during the initial drywell entry for a reactor shutdown (QlM16).
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).U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION Quad Cities Nuclear Power Station Unit1 05000254 NUMBER NUMBER 2003 001 00 3of4 (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 lIK] provides a Control Room alarm if particulate activity reaches a predetermined value. Also, the Drywell air temperature IM] is trended daily in accordance with COS 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.
A visual inspection was performed of the additional socket welds on the Unit 1 line and on the similar coupling welds on Unit 2. No additional weld defects were identified.7-2001)U.S. NUCLEAR REGULATORY COMMISSION (7-2001)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL REVISION Quad Cities Nuclear Power Station Unit 1 05000254 NUMBER NUMBER 2003 001 00 4 of 4 (If more space Is required, use addtonal copies of NRC Form 366AX17)
F.
PREVIOUS OCCURRENCES
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.