05000285/LER-2015-005, Regarding Reactor Coolant Leak at Reactor Coolant Pump Seal Due to Cyclic Fatigue

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Regarding Reactor Coolant Leak at Reactor Coolant Pump Seal Due to Cyclic Fatigue
ML15260B357
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 09/17/2015
From: Cortopassi L
Omaha Public Power District
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-15-0106 LER 15-005-00
Download: ML15260B357 (4)


LER-2015-005, Regarding Reactor Coolant Leak at Reactor Coolant Pump Seal Due to Cyclic Fatigue
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications

10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
2852015005R00 - NRC Website

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jjjjjjjj Omaha Public Power District 444 South 16" Street Mall Omaha, NE 68102-2247 LIC-15-0106 September 17, 2015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 10 CFR 50.73

Subject:

Licensee Event Report 2015-005, Revision 0, for the Fort Calhoun Station Please find attached Licensee Event Report 2015-005, Revision 0. This report is being submitted pursuant to 1 O CFR 50. 73(a)(2)(i)(B) and 50. 73(a)(2)(ii)(A). There are no new commitments being made in this letter.

If you should have any questions, please contact Brad Blome, Manager, Site Regulatory Assurance, at (402) 533-7270.

ouis P. Cortopassi Site Vice President and CNO LPC/epm Attachment c:

M. L. Dapas, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S.M. Schneider, NRC Senior Resident Inspector

NRC FORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (02-2014)

, the NRC may not conduct or sponsor, and a person Is not required to respond to, the information collection.

YEAR 2015

6. LER NUMBER I SEQUENTIAL I NUMBER 005 REV NO.

00 2

3. PAGE OF 3

Fort Calhoun Station (FCS) is a two-loop reactor coolant system (RCS) of Combustion Engineering design.

EVENT DESCRIPTION

The station started up from a refueling outage on June 10, 2015. On June 19, unidentified reactor coolant system (RCS) leakage was determined to be greater than 0.163 gpm and reactor coolant pump (RCP) seal pressure plots of the middle and upper seals of RCP RC-3A showed a steady decrease.

A containment entry made on July 8 identified a leak at RC-3A but station personnel only viewed the leak from the front. Station personnel did not access the back of the pump to identify the exact source of the leak. Based on the trended data and the walkdown it was believed that the leak was from the mechanical seal connections such as the shaft or 0-ring. On 7/15/15 an additional walkdown was performed and included visual inspection of the backside of RC-3A. Based on the walkdown video evidence and existing available data it was concluded that the failure location was from one of two mechanical seal 0-rings. An additional containment entry at power was performed on July 19, 2015.

This walkdown confirmed that RC-3A had no additional leakage sources and was confined to the one location on the back side of the pump.

On July 20, 2015, an increase in RCS leak rate required station personnel to perform a controlled shutdown of the reactor (Condition Report (CR) 2015-09023). The walkdown on July 21 (CR 2015-09130) determined that the location was not at the mechanical seals but rather at the middle seal cartridge inlet pressure tap as the result of a through-wall crack. The location of the crack was at the toe of the weld where the piping connects to the mechanical seal. The piping is a class 1 pressure boundary. Due to some initial confusion, station personnel did not immediately recognize this piping as part of the RCS pressure boundary.

At 1701 CDT on July 22, 2015, the station notified the NRG Headquarters Operations Office (HOO) of RCP seal leak as a crack per 10 CFR 50.72(b)(3)(ii)(A). The report was late due to the failure of station personnel to recognize this part of the seal as reactor coolant system boundary. This report is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) and 50.73(a)(2)(ii)(A).

CAUSE OF THE EVENT

Altran was contracted by Fort Calhoun Station to perform a failure analysis of the middle seal inlet pressure tap through-wall crack. Report 15-0456-TR-001 was completed to document all findings. The report was performed to determine the cause of the crack and identify any initiating contributors.

The conclusion of the report establishes the direct cause of the failure was low stress high cycle fatigue. The failure was noted to have started at the toe of the weld where the crack progressed inward from the outside diameter (OD) to the Inside diameter (ID). The report establishes that the NRC FORM 366 (02-2014)

6. LER NUMBER I SEQUENTIAL I NUMBER 005 REV NO.

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OF 3

geometry of the fillet weld resulted in a stress riser which when stressed in a cyclic fashion initiated the fatigue cracking.

The station assembled a team to review the issue and determine the root cause of the problem. The station has determined that the probable root cause was:

A design weakness resulted in the vibrations from RC-3A combined with the cantilevered pipe load caused cyclical stresses on the toe of a weld on the seal inlet pressure pipe tap. These stresses initiated a fatigue crack at the toe of the weld on the piping which subsequently propagated inward and progressed through the pipe wall causing the failure.

CORRECTIVE ACTIONS

Short Term Corrective Actions The station was shutdown using normal operating procedures and the seal package for RC-3A was replaced with a spare unit.

Long Term Corrective Actions The station will review design options to determine the appropriate resolution to the design issue with the seal package piping.

SAFETY SIGNIFICANCE

The leak occurred in a 3/4 inch schedule 80 pipe connected to a pressure sensor. The leak was well within the capacity of the charging system. This leak is bounded by station design analyses.

SAFETY SYSTEM FUNCTIONAL FAILURE This does not represent a safety system functional failure in accordance with NEI 99-02, Revision 7.

PREVIOUS EVENTS There have not been any previous incidents of cracking on N7500 reactor coolant pump seal piping at the station.

NRC FORM 366 (02-2014)