05000247/LER-2015-001

LER-2015-001, Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size that Results in Exceeding the Allowed Leakage Rate for Containment

Indian Point 2
Event date:	08-11-2015
Report date:	09-15-2016
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(v), Loss of Safety Function
2472015001R01 - NRC Website

LER 15-001-02 for Indian Point, Unit 2, Regarding Technical Specification (TS) Prohibited Condition Due to an Inoperable Containment Caused by a Service Water Pipe Leak with a Flaw Size that Results in Exceeding the Allowed Leakage Rate for Containment

ML17248A466
Person / Time
Site:	Indian Point
Issue date:	08/29/2017
From:	Vitale A J Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-107 LER 15-001-02
Download: ML17248A466 (7)

Note: The Energy Industry Identification System Codes are identified within the brackets fl.

DESCRIPTION OF EVENT

On August 11, 2015, at approximately 16:45 hours, during operator investigations inside the reactor containment building {NH}, a through wall leak was discovered on the 24 Fan Cooler Unit (FCU) {FCU} motor cooler service water (SW) {BI} return line. The leak was in a 2 inch 90-10 Copper-Nickel pipe {PSP}, line #495-SWN-NF near a brazed joint upstream of containment penetration SS at elevation 76 feet. The condition was recorded in the Indian Point Energy Center (IPEC) Corrective Action Program. (CAP) in Condition Report CR-IP2-2015-03550. NDE by UT was performed to characterize the flaw and the results evaluated using the structural margins provided in ASME Code Case N-513-3.

The leak was located within the ASME Section XI Code, Class 3 boundary and estimated to be approximately 2 gpm. Since the pipe flaw was through wall and was located within the ASME Section XI boundary, it exceeded the flaw allowable limits provided per IWC- 3000. Since the leaking defect was determined to be within the structural limits of the ASME Code Case N-513-3, the condition was determined to have no adverse impact on SW cooling safety function or on the structural integrity of the system. There was no visual indication of weld or base metal degradation at the affected pipe section, and there is no evidence of leakage at any other location on this weld or elsewhere on the piping adjacent to it. The leakage, if not contained, would have eventually drained into the containment sump. The impact of the pipe flaw was evaluated for containment free volume in-leakage and the leak is not expected to exceed the limit. The pipe is considered a closed loop system inside containment and required to meet containment integrity. An engineering evaluation was performed to determine the potential air leakage out of containment based on observed SW leakage into containment. This evaluation concluded that the leaking defect could result in post-LOCA air leakage out of containment in excess of that allowed by Technical Specification (TS) Surveillance Requirement (SR) 3.6.1.1 (Containment) which requires leakage rates to comply with 10 CFR 50, Appendix J.

The leak in 2 inch line #495-SWN-NF at the affected location is downstream of 24 FCU motor cooler which can be supplied with SW via either 18 inch line #408 (4-5-6 SW Header) or 18 inch line #409 (1-2-3 SW Header). At the time of discovery, the SW System (SWS) was aligned with the 4-5-6 SW Header as the Essential Header for Modes 1-4 Operations per Technical Specification (TS) 3.7.8 (Service Water System). The 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 design ensures a continuous flow of cooling water to those systems and components necessary for plant safety during normal operation and under abnormal or accident conditions. The SWS consists of two separate, 100% capacity, safety related cooling water headers. Each header is supplied by 3 pumps each having its own strainer, with SWS heat loads designated as either essential or non-essential. The essential SWS heat loads are those which must be supplied with cooling water immediately in the event of a Loss of Cooling Accident (LOCA) and/or Loss of Offsite Power (LOOP). The essential SWS heat loads can be cooled by any two of the three SW pumps on the essential header. Either of the two SWS headers can be aligned to supply the essential heat loads or the non-essential SWS heat loads. The design pressure and temperature of the SWS is 150 psig and 160 degrees F.

The function of line #495 is to return the SW that was used to cool the 24 FCU motor out of containment and discharge it to the discharge canal.

FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)

Cause of Event

The direct cause was corrosion. Corrosion of the copper-nickel piping resulted in a through wall pipe defect and leakage of SW into the reactor containment. The leak was estimated to be approximately 2 gpm. Based on the UT wall thickness measurements the leak occurred at a degraded area approximately 0.5 inch long in both the axial and circumferential directions. Based on the Code Case N-513-3 evaluation, the allowable flaw length in the axial and circumferential direction was calculated to be 1.47 inches and 1.10 inches respectively. Using a calculated corrosion rate of 4.7 mil (0.0047 inches) per year based on measured conditions, the predicted flaw length by the next refueling outage in the spring of 2016 is estimated to be approximately 0.51 inches in both the axial and in the circumferential directions, which are within the allowable flaw lengths. Based on these results, the affected piping is structurally acceptable consistent with the requirements of ASME code Case N-513-3. The specific corrosion type was most likely pitting corrosion or flow-assisted erosion-corrosion.

The apparent cause was the length of time to implement a modification to replace the FCU motor cooler copper-nickel piping identified in 2009 per the SW mitigation strategy. The Site Integrated Planning Database (SIPD)-1049 was not appropriately funded.

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:

• A leak-limiting engineered clam-shell type clamp was applied to the pipe flaw until a code repair could be completed.

• The clamp was monitored daily by a special operator log for any signs of increased .

leakage until a code repair could be completed.

• UT monitoring was performed every 90 days until the pipe was repaired.

• The pipe and affected elbow was replaced in accordance with the requirements of ASME Section XI Code during the spring refueling outage in 2016.

• The removed pipe/elbow was sent out to a vendor (LPI) for a metallurgical analysis to determine/confirm the specific cause.

• SIPD-1049 will be processed for funding for the project to replace the SW piping supply and return from the FCU motor coolers with upgraded materials.

FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)

Event Analysis

The event is reportable under 10 CFR 50.73 (a) (2) (i) (B) and 10 CFR 50.73 (a) (2)(v). The licensee shall report any operation or condition which was prohibited by the plant's TS. This condition meets the reporting criteria because TS 3.6.1 Containment Operability was not met. Because the FCUs are utilized to maintain normal containment temperature within accident analysis input limits and for accident mitigation, SW flow is not isolated following a DBA. SW piping supply and discharge for the FCUs is considered to be a closed system in containment or an extension of the containment boundary. Consequently, defects discovered within this piping may adversely affect containment integrity, and the ability to control release of radioactive materials.

The isolation valve for the faulted FCU SW piping was operable but the line was not isolated therefore the applicable TS not entered.

Initial operability determined the leak was within the Code Case Limits, that the inventory loss was not significant with no impact on other safety related structures, systems and components. Initially TS 5.5.14 requirements was not evaluated but subsequently it was determined the 10 CFR 50, Appendix J requirements were more stringent. The SW pipe flaw leakage was evaluated for containment in-leakage for potential flooding and out leakage for containment integrity per TS 5.5.14 and 10 CFR 50, Appendix J. TS 5.5.14.e requires that SW in-leakage into containment be limited to less than 0.36 gpm per FCU. Prior testing provided measured leakage that was less than the TS limit. Containment out leakage is required to be in accordance with TS Surveillance Requirement (SR) 3.6.1.1 whose leakage rate requirements comply with 10 CFR 50, Appendix J, Option B. The current total Appendix J leakage is 51,616.35 cc/min, therefore the total air leakage through the pipe defect must be limited to no more than 77,677.65 cc/min. A calculation based on a more limiting value of 70,000 cc/min of containment air leakage through the defect at 47 psig containment pressure and 7.41 psig of pipe internal pressure during LOCA conditions was determined to be equivalent to approximately 0.024 gpm of SW leakage out of the defect under current operating pressure (approximately 14.47 psig pipe internal pressure). Since TS 3.6.1 requires the containment to be operable in Modes 1-4 and TS SR 3.6.1.1 requires leakage rate requirements comply with 10 CFR 50, Appendix J, Option B and the estimated pipe flaw leak rate (2 gpm) exceeded the calculated equivalent containment air outleakage, TS 3.6.1 was not met. The application of an engineered clamp on the pipe flaw affected pipe section is structurally acceptable while degraded and the pipe considered operable.

The condition was also a safety system functional failure reportable under 10 CFR 50.73(a)(2)(v). The licensee shall report any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (C) Control the release of radioactive material. This condition meets the reporting criteria because TS 3.6.1 Containment Operability was not met. Because the FCUs are utilized to maintain normal containment temperature within accident analysis input limits and for accident mitigation, SW flow is not isolated following a Design Basis Accident (DBA). SW piping supply and discharge for the FCUs is considered to be a closed system in containment or an extension of the containment boundary. Defects discovered within this piping may adversely affect containment integrity, and the ability to control release of radioactive materials. As a result of revised reportability guidance provided by Entergy fleet experience, the inoperability of single train systems (e.g., containment barrier) are to be considered as a safety system functional failure (SSFF). This event was not at the time considered a SSFF therefore an event notification under 10CFR50.72(b)(3)(v) was not provided.

Past Similar Events

A review was performed of the past three years of Licensee Event Reports (LERs) for events reporting a TS violation due to inoperable SW piping caused by leaks and one LER was identified. This LER is a result of an extent of condition review. LER-2013-004 reported pin hole leaks in Code Class 3 SW piping elbows for series 300 stainless steel. The pin hole leaks were due to pitting corrosion. The cause of the event reported in LER-2013-004 was not the same as this event as the piping material was different (copper-nickel vs stainless steel), was in an elbow and the cause was failure to follow procedure EN-DC-336 (Plant Health Committee).

Safety Significance

This event had no effect on the health and safety of the public. There were no actual safety consequences for the event because there were no accidents or events during the degraded condition.

There were no significant potential safety consequences of this event. The leakage from the affected SW pipe was within the capability of the SW system to provide adequate SW flow to SW loads. The degraded piping was on the discharge of the FCU motor therefore any failure would not prevent the SW cooling function. Current analysis for SW pipe failures are postulated to be limited to small through-wall leakage flaws as SW is defined as a moderate energy fluid system. The SW leak would eventually drain to the containment sump. The containment sumps have pumps with sufficient capacity to remove excessive leakage.

The impact of the pipe flaw was evaluated for containment free volume in-leakage per the limits of TRM 3.4.D. The leak did not and is not expected to exceed the TRM 3.4.D limit. The pipe leak was just upstream of outboard containment isolation valve SWN-71- 4A.

SW effluent is monitored by radiation monitors R-46 and R-56 prior to discharge. If radiation is detected, each FCU heat exchanger can be individually sampled to determine the leaking unit. The SW for the 24 FCU can be isolated to prevent radioactive effluent releases. The Containment Spray System and Containment Fan Cooler System are Engineered Safety Feature systems designed to ensure that the heat removal capability required during the post-accident period can be attained. The CSS and the Containment FCU System provide redundant methods to limit and maintain post accident conditions to less than the containment design value. Five FCUs alone, or 3 FCUs and 1 CSP, or no FCUs and 2 CSPs possess this capability. The configuration with one CS train and two FCU trains is the configuration available following the loss of any safeguards power train (e.g., diesel failure). Accident analysis assumptions regarding containment air cooling and iodine removal are met, by one CS train and any two FCU trains (i.e., at least three FCUs). The Containment FCU System consisting of five 20 percent capacity FCUs and the CSS consisting of two 50% trains are divided into trains based on the safeguards power train which supports them. During the period of time the FCU SW pipe leak was being addressed there was minimum safeguards capability available. FCU 24 is associated with Fan Cooler Train 2A/3A which consists of the 24' FCU and the 23 FCU. An evaluation of the leak concluded that it did not result in any structural, flooding, or spraying condition that would adversely impact the capability of SSCs to perform their safety function.