05000247/LER-2016-010

LER-2016-000, Safety System Functional Failure Due to an Inoperable Containment Caused by a Through Wall Defect in a Service Water Supply Pipe Elbow to the 24, Fan Cooler Unit

Docket Number ,
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
Initial Reporting
ENS 52388	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
2472016000R01 - NRC Website

LER 16-010-01 for Indian Point 2 Regarding Safety System Functional Failure Due to an Inoperable Containment Caused by a Through Wall Defect in a Service Water Supply Pipe Elbow to the 24 Fan Cooler Unit

ML17069A170
Person / Time
Site:	Indian Point, Docket Number ,
Issue date:	02/28/2017
From:	Vitale A J Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-008 LER 16-010-01
Download: ML17069A170 (7)

comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to Note: The Energy' Industry Identification System Codes are identified within the brackets {}.

DESCRIPTION OF EVENT

On November 21, 2016, while at 100 percent reactor power, Operations investigated an increase in level rise in the Waste Holdup Tank (WHUT) {WD} and identified a corresponding rise in Containment {NH} sump level. At 16:50 hours, Entered Technical Requirements Manual (TRM) 3.4.D (Containment Free Volume Leakage) Condition A (Total leakage into the containment free volume greater than 10 gpm). At 16:55 hours, entered procedure 2-A0P-Flood-1 (Flooding) due to calculated leakage into containment of approximately 15 gpm. At 17:30 hours, a Containment

• entry was made to investigate the source of the sump level rise. The Control Room was notified at 17:38 hourS, the investigation by Operations identified the source of the leak was a through wall defect in a Service Water {BI} supply pipe elbow {PSF} to the 24 Fan Cooler Unit {FCU}. Entered Technical Specification (TS) 3.6.1 (Containment)., Condition A (Containment Inoperable) due to the possibility of a loss of containment {NH} integrity. The 24 FCU was isolated and removed from service at approximately 17:38 hours. Entered TS 3.6.6 (Containment. Spray System and Containment Fan Cooler System), Condition C (One Containment FCU Train Inoperable). After completing isolation of the 24 FCU SW, TS 3.6.1 was exited. As SW is credited as a containment boundary the defect in the SW pipe was determined to be a loss of safety function requiring an 8-hour non-emergency notification to the NRC. A non-emergency notification was made to the NRC for a safety system functional failure under 10CFR50.72(b)(3)(v) by Event Notification number 52388 at 21:22 hours. The leak was recorded in Indian Point Energy Center (IPEC) corrective action program (CAP) as CR- IP2-2016-06934 and CR-1P2-2016-07271.

The SW System (SWS) {BI} 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 to include pump strainers, 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 leak was a defect in a pipe fitting that is within the ASME Section XI Code ISI Class 3 boundary. The pipe fitting leak is in a moderate energy ASME ISI Code Class 3, nuclear safety related piping system. Because the defect in the elbow was through wall and was located within the ASME Section XI boundary, it exceeded the flaw allowable limits provided per IWD-3000. Since the through wall defect was located in a welded fitting (elbow), the ASME Code Case N-513-3 could not be applied because it excludes socket welds and pipe fittings.

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CONTINUATION SHEET

The pipe elbow for the SW supply to the 24 FCU is epoxy on approximately the 76 foot elevation in containment.

next to weld of the elbow to SW supply pipe.

coated carbon steel located The through wall flaw was The Containment Fan Cooler system consists of five 20 percent capacity FCUs located inside containment. These FCUs are used for both normal and post-accident cooling of the containment atmosphere. Each FCU consists of a motor,fan, cooling coils, dampers, duct distribution system and instrumentation and controls. SW is supplied to the cooling coils of each FCU to perform the heat removal function. During normal operation, SW is supplied to all five FCUs and one or more FCUs may be operated for containment cooling. It is necessary to limit the ambient containment air temperature during normal operation to less than the limit specified in TS 3.6.5 (Containment Air Temperature).

An extent of condition review included a best effort visual inspection of all FCUs.

There,was no evidence of 'additional leakage at any other place in the 24 FCU supply or return line or at any other location in the other four FCU lines 3 inch SW supply or return lines. The potential exists for a similar condition of lining degradation on

• the remaining' spool pieces in the 24 FCU that were previously Enecon repaired..

Unit 3 is not applicable to this condition as the piping material is not similar.

Unit 3 utilizes 3 inch Schedule 40, 904L alloy stainless steel piping and does not have interior epoxy lining..

CAUSE OF EVENT

The direct cause was failure of the interior Enecon (epoxy) coating allowing brackish river water to corrode the carbon steel pipe fitting. The pipe elbow with the flaw resulted in containment out leakage in excess of 10CFR50, Appendix J limits. The root cause was maintenance coating'procedure (0-SYS-409-GEN) requirements for post-coating inspections were inadequate. A defect or holiday in the applied coating remained undetected which eventually resulted in a coating functional failure. The coating failure in turn allowed corrosion to initiate and propagate through-wall by allowing brackish river water to contact the interior of the carbon steel pipe fitting (elbow).

CORRECTIVE ACTIONS

The following corrective actions have been or will be performed under the Corrective Action Program (CAP) to address the causes of this event:

• The defective elbow was removed, code'weld repaired, recoated on the interior and reinstalled.

• Maintenance procedure 0-SYS-409-GEN will be revised to include a mandate for detailed inspections and/or testing of the surface preparation and applied coatings to ensure proper coverage and adhesion. Requirements will be added to document the specific components coated and extent of the coatings applied to each component. A requirement will be included that any changes in the scope of the components coated are to be documented and concurred with by engineering.

comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by intemet e-mail to to impose an information collection does not display &currently valid OMB control number, the NRC may Indian Point 2 05000-247

• The IPEC Site Specific Curricula for GL 89-13 Program Inspections (Qualification Card) will be updated to include specific requirements for coating and lining inspections for FCU piping.

• A detailed visual inspection will'be performed of all 3 inch FCU SW piping elbow coatings during the FCUs scheduled Preventive Maintenance.

• Procedure SEP-SW-IPC-001 will be revised to include a requirement in the GL 89-13 Program to conduct and document a 100 percent internal lining visual inspection of all 3 inch FCU piping spool pieces, when removed during future FCU cooling coil PMs.

EVENT ANALYSIS

The event, is reportable under 10 CFR 50.73(a)(2)(v)(C). 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 meta The pipe flaw leakage was determined to be greater than the 10 CFR 50, Appendix J allowable leak rate. TS 3.6.1 (Containment) requires the containment to be operable in Modes 1-4. - TS Surveillance Requirement (SR) 3.6.1.1 requires visual examinations and leakage rate testing in accordance with the containment Leakage Rate Testing Program specified in TS 5.5.15. SR 3.6.1.1 leakage rate requirements comply with 10 CFR 50, Appendix J, Option B. As SW is required in an accident, the SW to the FCU would not be isolated in DBA and the piping credited as a closed system inside containment for containment integrity.

ConsequentlY, defects discovered within the FCU SW piping may adversely affect containment integrity and the ability to control releases of radioactive materkal.

The condition also meets the reporting criteria of 10 CFR 50.73(a)(2)(i)(B). The licensee shall report any operation or condition which was prohibited by the plant's TS. Dufing the previous period of operation for an unknown period of time the SW pipe contained a through wall leak that did not meet code requirements. This previously unrecognized condition required entry into TS 3.7.8 and corrective actions implemented to return.the pipe to operable. Failure to comply with the TS LCO and perform required actions is a TS prohibited condition.

PAST SIMILAR EVENTS

A review of the past three years of Licensee Event Reports (LERs) for, events that involved containment integrity due to flawed piping credited as a closed system inside containment identified two LERs. LER-2015-001-01 reported a Technical Specification prohibited condition and a Safety System Functional Failure (SSFF) due to a through wall leak on the 24 FCU motor cooler SW return line that results in exceeding the allowable leakage rate for containment. This LER is similar as the SW pipe defect was in piping credited as a closed system for containment integrity.

However, the pipe material (copper-nickel) and function (motor cooler return) are different. The direct cause was similar (pitting corrosion) but the apparent cause was different (length of time to replace the copper-nickel piping) as this LER concerns epoxy coated carbon steel. LER-2015-004 reported a SSFF due to an inoperable containment caused by a flaw on an elbow on the 24 FCU SW motor cooler return line.

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SAFETY SIGNIFICANCE

This condition 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. 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 and instrumentation to alert operators to a degraded condition.

The containment consists of the concrete reactor building, its steel liner, and the penetrations through the structure. The containment building is designed to contain radioactive material that might be released, from the reactor following a design basis accident (DBA). The containment building steel liner and its penetrations establish the leakage limiting boundary of the containment.

Maintaining the containment operable limits the leakage of fission product radioactivity from the containment to the environment. The DBA analysis assumes that the containment is operable such that., for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage.

The containment was designed with an allowable leakage rate of 0.1 percent of containment air weight per day. Containment isolation valves form a part of the containment pressure boundary. -Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analysis. One of these barriers may be a closed system such as the SW piping for the FCUs. The only time containment integrity can be affected is post accident when the FCUs safety function is being performed and SW pressure for the FCU cooling piping and coils may fall below peak accident pressure. 'Mitigation of radiation release by the degraded SW pipe pathway can be by use of radiation monitors R-46 and R-53 which monitor containment fan cooling water for radiation indicative of a leak from the containment atmosphere into the cooling water. 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.

During the time the FCU SW piping was degraded there was no leakage out of containment.