05000254/LER-2013-003
| Quad Cities Nuclear Power Station Unit 1 | |
| Event date: | 03-26-2013 |
|---|---|
| Report date: | 05-28-2013 |
| Initial Reporting | |
| ENS 48853 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| 2542013003R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EllS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
Pressure Boundary Leakage on Class 2 Reactor Head Vent Piping Identified During Refueling Outage VT-2 Examination
A. CONDITION PRIOR TO EVENT
Unit: 1 Event Date: March 26, 2013 Event Time: 1541 hours0.0178 days <br />0.428 hours <br />0.00255 weeks <br />5.863505e-4 months <br /> Reactor Mode: 4 Mode Name: Cold Shutdown Power Level: 0%
B. DESCRIPTION OF EVENT
On March 26, 2013, at 1541 hrs, while Unit 1 was in cold shutdown for refueling outage Q1R22, a 20 drop per minute (dpm) leak was identified at the insulation [ISL] surrounding the reactor vessel [AD] head vent piping [SB]. The leak was discovered during the Reactor Pressure Vessel (RPV) Class 1 pressure boundary system leakage test that is performed each refueling outage in accordance with ASME Section XI, IWB-2500, and station procedures. During the test, the vessel was water-solid and pressurized to 1005 psig.
Upon discovering the leak, the qualified VT-2 inspector removed the insulation to allow a more detailed inspection.
The leak was confirmed to be originating in the RPV head vent piping from a socket weld between the piping and a 90-degree elbow. A leak at this location is considered a through-wall failure of the primary coolant [AD] system. At 1645 hrs, a second VT-2 inspector verified that the leak appeared to be through-wall and the RPV leak test was terminated. The reactor vessel was depressurized and vented to atmosphere at 1826 hrs the same day.
On March 27, 2013, at approximately 0245 hrs a Level 3 Non-Destructive Examination (NDE) Inspector confirmed the presence of a round "void" (indicative of a pin-hole) approximately 0.15-inch deep, at the location of the leakage. The void was evident by visual observations (VT-1). The head vent piping was unbolted and transferred to the Refuel floor. The weld was removed by grinding and repaired with an EPRI 2:1 weld leg profile, including post weld NDE.
The repair was completed on March 27, 2013 and the post-repair leakage test was performed at 1445 hrs on March 27, 2013. Successful system leak testing validated the integrity of the repair and the RPV attached head vent piping, which is a Class 2 pressure boundary.
On March 27, 2013, ENS Notification#48853 was made in accordance with 10 CFR 50.72 (b)(3)(ii)(A). This event has been classified as a Maintenance Rule Functional Failure for the Main Steam system.
Given the impact on the reactor coolant pressure boundary, this report is submitted in accordance with the requirements of 10 CFR 50.73 (a)(2)(ii)(A), which requires the reporting of any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.
C. CAUSE OF EVENT
The apparent cause of the RPV Class 2 head vent line weld pin-hole leak was determined to be a defect in the socket weld, attributed to porosity and/or slag that originated from initial construction in 1970, prior to the initial start-up of Unit 1.
D. SAFETY ANALYSIS
System Design The purpose of the RPV and its appurtenances, such as this section of the non-isolable Class 2 head vent pipe, is to retain the reactor core coolant-moderator within the RPV and to serve as a high integrity barrier against leakage of steam used for power production and leakage of radioactive materials to the drywell [NH] during all modes of plant operation. The RPV head vent line provides a connection from the RPV upper steam dome volume to a vent path to the drywell sumps [WK] during shutdown conditions and to the main steam [SB] lines to vent off non-condensable gasses during power operation.
The FW14 weld is located on the RPV head vent line and during plant operation is exposed to the steam section of the RPV.
Safety Impact The safety significance of this event was minimal given the leakage was very small, was found while the reactor was shutdown for refueling, and if this leak had occurred during plant operation, it would not have exceeded TS leakage limits for unidentified RCS leakage. There was no evidence this leak had occurred during the prior operating cycle.
Increases in leakage from these types of defects would be due to steam cutting over time and would likely grow slowly.
This associated failure mechanism is highly localized and is slow acting. The defect does not result in a crack that is likely to propagate around the weld. Therefore, it is highly unlikely to result is a sudden large leak or catastrophic failure. Periodic leak inspections are performed at least once per operating cycle; this frequency is adequate to identify these types of thru-wall leaks before they grow. Additionally, continuous RPV leak monitoring [IJ] systems (e.g., drywell sumps and air monitoring and sampling systems) are in place to detect primary system boundary leakage inside the drywell during power operation.
During power operation, this pipe section is exposed to a reactor steam environment, and not the water solid conditions of the Class 1 pressure boundary system leakage test that is performed each refueling outage in accordance with ASME Section XI, IWB-2500.
Had the 2-inch diameter FW14 weld catastrophically failed, such that there was a 2-inch diameter opening in this head vent piping as connected to the RPV, the consequences would have been minimal. The 2-inch diameter line break (0.02 sq ft in area) would be bounded by the reactor coolant system line breaks of up to 0.12 sq ft in area, as discussed in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR), which provides that the High Pressure Coolant Injection (HPCI) [BJ] system could supply sufficient coolant to depressurize the vessel and cool the core for line breaks of up to 0.12 sq ft in area (small break LOCA).
The as-found leak at the FW14 weld would have had no effect on the ability of the head vent line to remove condensable gasses from the RPV during power operation, or impact the capability of the RPV to be vented during cold shutdown conditions.
Risk Insights Prior to and during this event, the capability of the reactor head vent system was not lost. This condition did not create any actual plant or safety consequences since the Unit was not in an accident or transient condition during this period of time or prior operating cycle. The FW14 weld leak was discovered during cold shutdown system pressure testing which is designed to identify any RPV leakage before the unit is placed into power operation.
There is no makeup rate required to mitigate an RPV leakage of 20 dpm. The head vent line at the FW14 weld is a 2- inch pipe/fitting. Even a complete failure of the weld would only result in Small LOCA (SLOCA). The break diameter of 2 inches would be well within the capability of HPCI or one Residual Heat Removal [BO] (RHR) pump. The 20 dpm leak is negligible compared to any SLOCA.
Based on the above, the change in risk due to a 20 dpm leak is negligible. Therefore, considering the impact of this condition on the Plant Probabilistic Risk Assessment (PRA), the change in Core Damage Frequency (CDF) due to the observed leakage will be less than 1.0E-06/yr. In conclusion, the overall safety significance and impact on risk of this event was minimal.
E. CORRECTIVE ACTIONS
Immediate:
1. The degraded FW14 weld was removed and repaired to allow a new EPRI 2:1 leg weld to be installed.
2. Successful system leak testing validated the integrity of the repair and the RPV.
Follow-up:
1. The 2-inch socket welds on the head vent lines for both units will be inspected using dye penetrant (PT) methods during the next Unit 1 and Unit 2 refuel outages. Additional corrective actions will be identified as necessary as a result of these inspections.
2. The remaining 2-inch socket welds on the head vent lines for both units will be re-welded using EPRI 2:1 leg welds during the next Unit 1 and Unit 2 refuel outages. This action is in addition to performing the PT inspections above.
F. PREVIOUS OCCURRENCES
The station events database, LERs, INPO Consolidated Event System ICES (EPIX) were reviewed for similar events at Quad Cities Nuclear Power Station. This event was an RPV head vent line socket weld leak associated with an original weld defect. There was one other previous similar occurrence identified at Quad Cities Nuclear Power Station that involved an event of this type, and is discussed below.
- LER 254/2003-001-00, 7/21/03, Reactor Shutdown due to Reactor Head Vent Steam Leak Constituting Pressure Boundary Leakage (5/20/03) — This event was caused by 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 defect propagated due to long-term corrosion and possibly fatigue to form the through wall leak. Corrective actions from this event included: Removal of the failed weld, coupling, and two-foot section of pipe, and installing a new section of pipe and couplings. The new welds were visually inspected and liquid penetrant tested; and 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 at that time.
At the time of this 2003 LER event the causes and required corrective actions were limited to the knowledge of this isolated event. Since there are currently no approved volumetric exam techniques to interrogate these types of welds, available inspections are limited to surface NDE, including VT-1, PT, and MT. These surface exams will not identify internal flaws, unless the flaw has communicated to the surface either through corrosion, cracking, or an original defect. Since it is difficult to inspect these types of welds to enable predictions of impending leaks, the 2003 LER event causes and corrective actions were not able to directly contribute to preventing this 2013 LER event.
G. COMPONENT FAILURE DATA
The failed component was the FW14 weld on the RPV head vent line; this weld was field fabricated during initial plant construction. The RPV was constructed by B&W.
The FW14 weld is a 2 inch diameter pipe to 90-degree elbow weld on the RPV head vent line located in the reactor cavity and connected to the vessel head. Materials of construction are A106 carbon steel schedule 80 pipe and 3000 lb fittings.
This event has been reported to ICES as Report No. 305475.