05000286/LER-2011-003
| Indian Point 3 | |
| Event date: | 2-22-2011 |
|---|---|
| Report date: | 4-25-2011 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2862011003R00 - NRC Website | |
Note: The Energy Industry Identification System Codes are identified within the brackets {}.
DESCRIPTION OF EVENT
On February 22, 2011, at approximately 10:30 hours, while at 100% steady state reactor power, with the 1-2-3 service water (SW) header as the essential SW header {BI}, the Control Room {NA} received notification of flooding in the south SW valve pit.
Subsequent investigation by a Nuclear Plant Operator (NPO) determined there was approximately three feet of water in the valve pit with leakage in the area of valves SWN-6 and SWN-7 {HCV}. Valves SWN-6 and SWN-7 are the Conventional Essential Header Discharge isolation valves. Operations also noted the 1-2-3 SW header pressure had dropped approximately 3 psig from previous readings. Upon further investigation, the leak was estimated to be approximately 150 gpm and not isolable. At 12:03 hours, Technical Specification (TS) 3.7.9 (Service Water System) Condition E was entered for SW System (SWS) piping and valves inoperable for reasons other than Condition A, B, C, or D with no loss of safety function. Additional evaluations of the condition concluded there was a loss of safety function. The south SW valve pit contains SW valves SWN-FCV-1111, SWN-FCV-1112 {FCV}, SWN-6 and SWN-7 and is located outside the Turbine Building {NM}. Flooding due to the pipe leak would have prevented operators from closing these valves. Access to these valves is required in Standard Operating Procedures and Emergency Operating Procedures (EOP), specifically EOP 3-ES-1.3 (Transfer to Cold Leg Recirculation). As a result of a design basis accident (DBA), EOP procedure E-0 (Reactor Trip with Safety Injection) will be entered and will direct RO-1 (BOP Operator Actions During Use of EOPs) to be performed. RO-1 will direct closure of SWN-FCV-1111, SWN-FCV-1112, SWN-6 and SWN-7. It was concluded this condition was a safety system functional failure because of the requirement to start a non-essential SW {KG} pump which could go to run out due to the failure to isolate conventional SW loads. The condition was recorded in the Indian Point Energy Center (IPEC) Corrective Action Program (CAP) as Condition Report CR-IP3-2011-00680.
Prior to the event, insulators were in the south SW valve pit removing insulation on the affected line. After removing the last section of insulation a pop was heard and they noticed a leak. Operations was notified and an investigation was initiated.
The SW System (SWS) is designed to supply cooling water from the Hudson River to various heat loads in both the primary and secondary portions of the plant. The SWS needs to provide continuous cooling water flow to systems and components that are required during normal and accident conditions. There are six vertical centrifugal pumps {P} rated at 6000 gpm that supply two independent SW discharge headers (three pumps per header). Each header is connected to an independent 24 inch supply line and either line can be used to supply the essential loads while the other line feeds the non-essential loads. The essential SW loads are those that are required immediately in the event of a DBA. A 10 inch and a 16 inch supply line provide conventional SW loads.
The Conventional Non-Essential SW header receives SW through valve FCV-1111 from SW pumps 34, 35 and 36 or through valve FCV-1112 from SW pumps 31, 32, 33. Valves FCV- 1111 and FCV-1112 are manual butterfly flow control valves located in the south valve pit. Water is supplied to the Conventional Essential header through manual butterfly valve SWN-6 on 10 inch SW line #1222 from SW header #408 supplied by the 34, 35, 36 SW pumps, or through manual butterfly valve SWN-7 on 10 inch header #1221 from SW header #409 supplied by SW pumps 31, 32, 33.
A hole of approximately 3/4 inch was identified on a vertical portion of the 10 inch SW line #1222 downstream of valves SWN-6 and SWN-7 in the combined 10 inch Conventional Essential SW header. Ultrasonic Testing (UT) measurements of the pipe were evaluated and determined to be structurally acceptable.
Cause of Event
The direct cause of the leak was a 3/4 inch hole in the 10 inch Conventional Essential SW header line #1222 downstream of valve SWN-6 due to a through wall leak. The likely cause of the through wall hole was corrosion due to inadequate coating of the interior of the carbon steel pipe during a previous pipe repair.
The root cause was an inadequate installation plan and repair of a flaw identified in 1992. A review of work history for this portion of SW pipe identified a Work Order for line #1222 in the location of the leak that removed a temporary repair patch at a weld on the downstream flange of valve SWN-6 and installed a permanent repair. The method of repair chosen was a weld insert requiring a portion of the pipe to be cut out and a new piece installed. A WaterPlug epoxy coating was used to coat the inside of the pipe where the concrete lining was removed for the repair. The weld insert method did not take into account the difficulty of coating the interior of the carbon steel pipe once the cement lining was removed. The WO that was planned for the repair in 1992 could not be properly implemented and resulted in inadequately coated carbon steel piping.
Given the limited access to the pipe internals and the amount of pipe that would need to be coated after welding, the repair method should have been a replacement of the affected section of pipe.
Corrective Actions
The following corrective actions have been or will be performed under Entergy's Corrective Action Program to address the cause and prevent recurrence:
- Installation of a modified pipe clamp and UT readings to determine pipe operability
- The effected pipe section in line #1222 was replaced in spring 2011 refueling outage.
- An inspection of the removed pipe section verified the failure mechanism.
- An engineering guideline will be developed providing direction on how SW leak repairs should be performed.
- The Generic letter 89-13 program will be revised to prioritize inspection frequencies of SW pipe welds and include SW lines 1221 and 1222.
- Inspections of SW piping in the Unit 2 SW valve pit and the Unit 3 north SW valve pit will be performed.
Event Analysis
The event is reportable under 10CFR50.73(a)(2)(i)(A). The licensee shall report the completion of any plant shutdown required by the plant's Technical Specification. At 12:03 hours, TS 3.7.9 Condition E was entered for SWS piping and valves inoperable for reasons other than Condition A, B, C, or D with no loss of safety function. After further assessment of the condition, it was concluded at 14:25 hours, that the condition would prevent the credited method of isolating the SW loads in the event of a Design Basis Accident requiring cold leg recirculation due to the inaccessibility to the valve pit for valve operation. As a result of this determination TS 3.7.9 Condition E was no longer applicable and the plant was in a condition not specified in the TS. In accordance with TS 3.0.3 when an associated action is not provided, action shall be initiated within one hour to place the unit in Mode 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, Mode 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> and Mode 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />. TS 3.0.3 was entered at 14:25 hours and the shutdown commenced at 15:25 hours. At 18:38 hours, a manual reactor trip was initiated and the plant entered mode 3 (Hot Shutdown). Entry into TS 3.0.3 and completion of any plant shutdown required by the TS is reportable under 10CFR50.73(a)(2)(i)(A) and 10CFR50.73(a)(2)(i)(B).
The condition was also a safety system functional failure reportable under 10CFR50.73(a)(2)(v)(D) as the condition would have prevented access to valves in the valve pit used for accident mitigation. Failure to provide applicable valve manipulations could have resulted in inoperability of a non-essential SW pump required for accident mitigation.
Past Similar Events
A review was performed of the past three years of Licensee Event Reports (LERs) for events that involved a TS required shutdown or safety system functional failure. There were no LERs identified that were similar to this event. LER-2010-002 reported a manual reactor trip as a result of a SW leak in the exciter. However, the shutdown was not required by TS and was not a SSFF. CAs for that event would not have prevented this event as the causes were different.
Safety Significance
This event had no significant effect on the health and safety of the public. There were no actual safety consequences for the event because there were no accidents or transients during the time of the event. As an alternative to operating valves SWN-6 and SWN-7, FCV-1111 and FCV-1112 in the SW pit there are other valves that can be operated to isolate the conventional loads and prevent SW pump run out. In addition, there are three SW pumps on the designated non essential SW header. Therefore, there are remaining SW pump capabilities as backup to the SW pump designated for use in DBA recovery. A risk impact was performed of the flooding with an exposure time of 31.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (discovery on February 22, 2011 at 10:30 hours to operability declared on February 23, 2011, at 17:55 hours). Assuming operators would have been unable to enter the SW valve pit to operate isolation valves, the Core Damage Frequency (CDF) increased to 2.7E-5 per year which results in an incremental CDF (ICDF) of 1.24E-5 per year.
Considering the event for a 31.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> period of unavailability, results in an incremental core damage probability (ICDP) of 4.46E-8. This impact is not considered significant.