05000247/LER-2014-002
| Indian Point 2 | |
| Event date: | 02-24-2014 |
|---|---|
| Report date: | 04-23-2014 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii) |
| 2472014002R00 - NRC Website | |
Note: The Energy Industry Identification System Codes are identified within the brackets {}.
DESCRIPTION OF EVENT
On February 24, 2014, during initial Containment walkdowns after shutdown for a refueling outage, Operations identified a steam leak on 23 Steam Generator (SG) (AB} drain line valve MS-68. Valve MS-68 is a normally closed valve on a one inch drain pipe from the 23 SG shell side to a two inch SG Blowdown (SGBD) pipe that is routed to the blowdown tank. The one inch SG drain pipe and valve MS-68 are safety related, ASME Code Class 2 High Energy (HE) and Seismic Class 1 components. Engineering assessment of the condition determined the leak could not be isolated and Ultrasonic Testing (UT) could not be performed to determine the extent of the defect. The valve and associated piping are a pressure boundary for the SG and because there is no ASME Code method to evaluate the structural integrity of the through wall defect, engineering concluded the valve was inoperable. Operations declared the 23 SG inoperable but the plant was in Mode 5 (Cold Shutdown) and one SG inoperable did not impact Technical Specification (TS) 3.4.7 [Reactor Coolant System (RCS) Loops-Mode 5, Loops Filled]. TS 3.4.7 requires one Residual Heat Removal (RHR) loop to be operable and in operation, and either the non-operating RHR loop to be operable or the secondary side water level of at least two SGs to be equal to or greater than 0% narrow range. The condition was recorded in the Indian Point Energy Center corrective action program (CAP) as CR-IP2- 2014-00975.
Leakage in Containment although unrecognized from MS-68 was previously identified in July 2013 as an increase in the fill rate of the Containment Sump and activity of the Fan Cooler Unit weirs. This condition was recorded in the IPEC CAP as CR-IP2-2013- 03207. Several Containment entries were made to identify the source of the leakage but without success. An assessment of operability concluded there was no inoperable condition. There were no indications of actual RCS leakage and Pressurizer level, Volume Control Tank level, charging pump flow, RCS pressure and Radiation Monitors remained steady and within normal parameters. Containment sump and all sump levels remained within normal parameters. The calculation of RCS leakage maintained operability for TS 3.4.13 (RCS Operational leakage) and sump leakage maintained functionality for Technical Requirements Manual (TRM) 3.4.D (Containment Free Volume Leakage). An ODMI and troubleshooting plan was prepared and implemented.
The primary function of the RCS is removal of the heat generated in the fuel due to the fission process and transfer of this heat, via the SGs, to the secondary plant. The reactor coolant is circulated through four loops connected in parallel to the reactor vessel, each containing a SG, and a reactor coolant pump. The SGs provide the heat sink for the RCS. If the requirements of the TS Limiting Condition for Operation (LCO) are not met, the required action is to reduce power and bring the plant to Mode 3 (Hot Shutdown). This action lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating design limits.
The Main Steam (MS) {SB} and Feedwater (FW) {SJ} lines and the shell side of the SG are considered an extension of the containment boundary (closed system inside containment) (Class 7 Containment Isolation). Valve MS-68 is on a one inch pipe from the SG shell side that connects to the SGBD line which is then routed through a Containment penetration to the Blowdown Flash Tank and Sampling System. Two containment isolation valves in series are provided on the SGBD line as it exits Containment. The SGBD system is primarily used in maintaining the secondary side water chemistry of the SGs but also provides water samples and a means to drain the shell sides for inspection/maintenance. MS-68 is a one inch diameter gate valve {V) manufactured by Vogt (V139}, Model SW-2805.
Cause of Event
The cause of the steam leak was a pin hole in the body of valve MS-68. The pin hole was likely a defect in the original valve casting which over time propagated through the valve wall. Leakage was estimated to be approximately 0.4 gpm at SG secondary side pressure of approximately 750 psig.
Corrective Actions
The following corrective actions Program to address the cause and
- The defective MS-68 valve was
Event Analysis
have been performed under Entergy's Corrective Action prevent recurrence:
removed and replaced with a new valve (WO#375380).
The event is reportable under 10CFR50.73(a)(2)(i)(B). The licensee shall report any operation or condition which was prohibited by the plant's TS. This condition meets the reporting criteria because during past operation the defect in the valve body of ASME Code Class 2 HE component MS-68 leaked resulting in an increase in the fill rate of the Containment Sump and increased Fan Cooler Unit weir activity as recorded in CR- IP2-2013-03207. As a result of this defect, the valve and drain line can allow steam/feedwater on the SG secondary shell side to leak into containment. The valve and associated piping are a pressure boundary for the SG and because there is no ASME Code method to evaluate the structural integrity of the through wall defect, it was concluded the valve was inoperable. As the leak could not be isolated and Ultrasonic Testing (UT) could not be performed to characterize the defect, Operations declared the 23 SG inoperable. TS 3.4.4 [Reactor Coolant System (RCS) Loops-Modes 1 and 2] requires four RCS loops to be operable and in operation during Modes 1 and 2. An operable RCS loop consists of an operable reactor coolant pump in operation providing forced flow for heat transport and an operable SG. Because the condition was determined to be the cause of leakage in the Containment during past operation at full power and was considered to result in an inoperable 23 SG, the requirements of TS 3.4.4 were not complied with and is therefore a TS prohibited condition.
The condition is not reportable in accordance with 10CFR50.73(a)(2)(ii), Any event or condition that resulted in (A) The condition of the power plant, including its principal safety barriers, being seriously degraded; or (B) the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety. Although there was no ASME Code method to evaluate the structural acceptability of the through wall defect, engineering judgment concluded the condition was structurally intact and did not represent a potential large loss to the SG and would not result in containment leakage in excess of that allowed by 10CFR, Appendix J under shutdown accident conditions.
Therefore, there were no principal safety barriers seriously degraded nor was there any condition that significantly degraded plant safety.
Past Similar Events
A review was performed of the past three years for Licensee Event Reports (LERs) reporting a TS prohibited condition due to an inoperable ASME Code Class 2 boundary component. No LERs were identified. LER-2013-004 reported a TS prohibited condition due to an inoperable service water header as a result of a pin hole leak in Code Class 3 Moderate Energy SW piping. LER-2012-003 reported through wall defects in two reactor coolant pressure boundary (RCPB) components: #1) Defect on the horizontal leak-off pipe of Pressure Control Valve PCV-455A Spray Inlet Stop valve 4152, #2) Defect in the socket weld of a 3/8 inch diameter tubing fitting down stream of pressure transmitter isolation valve 4138. Neither of these were Code Class 2 valve body defects.
Safety Significance
There were no actual safety consequences for the event because there were no significant failures in components that are credited as closed systems inside containment for the Containment Isolation System (CIS). The CIS provides the means of isolating the various pipes passing through the containment walls as required to prevent the release of radioactivity to the outside environment in the event of a Design Basis Accident (DBA). A CIS design assumption is a piping rupture outside the containment at the same time as a DBA (i.e., LOCA) is not considered credible, as the penetrating lines are Seismic Class I up to and including the second isolation barrier.
The MS, FW and SGBD system have redundant isolation capability to prevent any release of radioactivity to the environment in the event of a DBA.
Leakage through the valve defect was approximately 0.4 gpm at SG secondary side pressure of 750 psig. This amount of leakage would not have prevented the Auxiliary Feedwater System from providing its design cooling capability. TS surveillances provide a trend of leakage early before significant degradation. An early warning of unidentified leakage is provided by the systems that monitor containment atmosphere radioactivity, operation of the containment sump, sump level, Fan Cooler Unit weir condensate, and Containment temperature and dew point. Monitoring and periodic inspections identify leaks when they are small so that repairs can be performed to prevent further degradation.
Although there was no ASME code case or analytical means to evaluate the structural acceptability of the through wall defect, Engineering judgment concluded the condition was structurally intact and did not represent a potential large loss to the SG and would not result in containment leakage in excess of that allowed by 10CFR, Appendix J under shutdown accident conditions.