05000278/LER-2017-001

LER-2017-001, Reactor Pressure Boundary Leakage Due to Weld Failure in One-Inch Diameter Instrument Line

Peach Bottom Atomic Power Station Unit 3
Event date:	10-23-2017
Report date:	12-21-2017
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
Initial Reporting
ENS 53031	10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
2782017001R00 - NRC Website

LER 17-001-00 for Peach Bottom Atomic Power Station (PBAPS), Unit 3 Regarding Reactor Pressure Boundary Leakage Due to Weld Failure in One-Inch Diameter Instrument Line

ML17355A003
Person / Time
Site:	Peach Bottom
Issue date:	12/20/2017
From:	Navin P D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CCN: 17-106 LER 17-001-00
Download: ML17355A003 (4)

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LER NUMBER

2017 - 00 001 Unit Conditions Prior to Discovery of the Event Unit 3 was in Mode 3, Hot Shutdown, when the condition was discovered on 10/23/2017 at approximately 4:00 am. The unit entered Mode 3 at 1:08 am in preparation for the planned refueling outage. There were no structures, systems, or components out of service that contributed to this event.

Description of the Event

On 10/23/2017, with the Unit in Mode 3, at the beginning of a refueling outage, personnel entered the drywell to perform an inspection. At approximately 4:00 am, leakage was identified on a one-inch diameter instrument line socket (EIIS:PSF) weld. The instrument line is for a pressure transmitter from the 'B' Recirculation Pump (EIIS:AD) discharge. Because the leak was misting, the leakage rate could not be quantified. However, the reactor coolant system unidentified leakage prior to plant shutdown was 0.18 gpm.

This line is part of the ASME Class 1 primary coolant pressure boundary. Technical Specification (TS) 3.4.4 does not allow any pressure boundary leakage in Modes 1, 2 or 3. As a result, TS 3.4.4 Condition C was entered, which requires the plant to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The plant was in Mode 3 at the time of discovery and entered Mode 4, Cold Shutdown, at 12:54 pm on 10/23/2017 (approximately 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after time of discovery).

Prompt notification to the NRC was made in accordance with 10 CFR 50.72(b)(3)(ii)(A) on 10/23/2017 at 9:35 am (Event Notification #53031).

Analysis of the Event

The leak was determined to be from a crack in the weld connecting the pipe to a tee fitting. The crack extended approximately 60 degrees around the circumference of the weld. Although it could not be determined when the crack developed, it is assumed it existed while operating in Mode 1, prior to shutdown. During such time, the leakage would have been contributing to the total unidentified leakage of 0.18 gpm. This is well below the 5 gpm limit for unidentified leakage as stated in TS 3.4.4.

This event is being reported in accordance with the following:

1. 10 CFR 50.73(a)(2)(i)(B) — Conditions Prohibited by Technical Specifications - TS Limiting Condition for Operation (LCO) 3.4.4 requires there to be no RCS pressure boundary leakage while in Modes 1, 2 and 3. If leakage exists, TS 3.4.4 Condition C requires the unit to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Since the leak existed while in Mode 1 for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, a condition prohibited by TS existed.

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to Infocollects.Flesource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (31500104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. LER NUMBER

2017 - 00 001 2. 10 CFR 50.73(a)(2)(ii)(A) — Degradation of the RCS — Because of the RCS pressure boundary leakage, one of the principal safety barriers of the plant was degraded.

Cause of the Event

The section of the one-inch pipe containing the cracked socket weld was sent to an offsite laboratory for analysis. The pipe was sectioned and lack of fusion defects were identified at the root of the weld. The crack initiated at the location of these defects. The defects would have occurred when the weld was performed during a recirculation system pipe replacement in the late 1980's. It is unknown when the crack began, but normal vibration for the piping likely caused it to propagate to the surface of the weld, resulting in the identified leak.

Corrective Actions

The section of pipe and the associated fitting were replaced. Instrument lines connected to the suction and discharge lines for both of the recirculation pumps with similar configuration and subject to vibration were also replaced during the refueling outage. The new welds were performed with a 2:1 profile, which reduces their susceptibility to vibration-induced failures.

Previous Similar Occurrences A similar event occurred in September of 2005. A crack in a socket weld caused by a lack of fusion defect resulted in a 1 gpm leak in a one-inch equalizing line for a check valve on the 'A' Residual Heat Removal (RHR) injection line. The event is documented in LER 2005-003, dated 10/28/2005. Additional information on this previous occurrence is contained in the corrective action program.