05000387/LER-2015-009

LER-2015-009, Pressure Boundary Leakage From an Inadequate Weld Repair in Small Bore Pump Seal Vent Piping

Susquehanna Steam Electric Station Unit 1
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
Initial Reporting
ENS 51538	10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
3872015009R00 - NRC Website

LER 15-009-00 for Susquehanna, Unit 1, Regarding Pressure Boundary Leakage From an Inadequate Weld Repair in Small Bore Pump Seal Vent Piping

ML16011A415
Person / Time
Site:	Susquehanna
Issue date:	01/11/2016
From:	Franke J A Susquehanna, Talen Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-7419 LER 15-009-00
Download: ML16011A415 (5)

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 internet 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.

CONDITIONS PRIOR TO THE EVENT

Unit 1 — Mode 4, Unplanned shutdown, 0 Percent Other than the leaking weld itself there were no systems, structures, or components at the start of the event that contributed to the event. Prior to this condition on November 12, 2015, at 1132 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.30726e-4 months <br />, the Unit 1 reactor automatically scrammed due to one Main Steam Isolation Valve (MSIV) unanticipated closure causing a High Pressure Reactor Protection System trip. During the unplanned forced outage, a drywell entry and drywell leakage investigation were performed.

EVENT DESCRIPTION

On November 13, at 1745 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.639725e-4 months <br /> during drywell entry, a leak was reported on the "B" Reactor Recirculation (RXR) [EIIS System Identifier: AD] Pump [EIIS Component Identifier: P] Lower Seal Cavity Vent piping [EllS Component Identifier: PSP]. The leak was identified at the inboard pipe-to-union weld and required a weld repair prior to returning Unit 1 to service. The leaking weld was attached to the vendor supplied replacement pump seal chamber. The affected piping weld is for 3/4-inch piping, schedule 80 SA-479 TP304 or TP316, union 3000# SA-182 Gr F304 or F316. The affected piping had been in service for approximately 11 months following a previous repair of the weld at this location during the December 2014 forced outage of Unit 1, (e.g., LER 2014-011-00, issued February 11, 2015). A timeline of relevant events follows:

Mid-1990s April 2013 April 2014 December 2014 December 2014 through 2015 November 12, 201 Replacement RXR pump shafts and related components were procured from FLOWSERVE (formerly BW/IP). The replacement shafts were stored in the warehouse until installed (one on the 2A RXR pump in 2013 and the other on the 1B RXR pump in 2014). The weld discovered to have cracking and leakage in December 2014 was a vendor weld that existed at the time of procurement.

The 2A RXR pump shaft and related components were replaced with components procured form FLOWSERVE in the mid-1990s.

The 1B RXR Pump shaft and related components were replaced with components procured from FLOWSERVE in the mid-1990s.

Unit 1 initiation of a plant shutdown due to unidentified drywell leakage, discovery of weld cracking at this location, and its repair.

Unit 1 drywell leakage took a step increase from 0.1 to 0.2 gallons per minute (GPM) on December 28, 2014. Subsequently, gradual increases in excess drywell leakage were observed, (e.g., 0.2 to 0.6 GPM).

5 Unit 1 reactor scram due to 'B' inboard MSIV closure during surveillance testing, (unrelated to this condition).

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 internet 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.

November 13, 2015 Drywell leakage is identified again to be at the 1P401B Reactor Recirculation Pump Seal Connection No. 8.

November 21, 2015 A weld repair was again performed on the 1B reactor recirculation pump seal.

This condition was reported on November 13, 2015, at 0022 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />, in accordance with 10 CFR 50.72(b)(3)(ii)(A) as a principal safety barrier degradation (EN 51538). This License Event Report (LER) is written in accordance with 10 CFR 50.73(a)(ii)(A) and 10 CFR 50.73(a)(2)(i)(B) as a condition that resulted in a principal safety barrier degradation with evidence of reactor coolant pressure boundary leakage, which is a condition prohibited by Technical Specifications.

CAUSE OF THE EVENT

The cause of this event is a decision making process that did not consider the technical merits of prudent versus allowable repair options. More specifically, the previous December 2014 weld repair did not fully excavate the weld and remove the J-groove, and thereby eliminate the presence of the crack.

A lack of fusion between original vendor weld passes, or at the root of the weld at the interface with the remaining vendor weld, is also considered to be a causal factor, contributing to further weld cracking and resulting reactor coolant pressure boundary leakage. This lack of fusion remaining within in the vendor weld, not in the December 2014 weld repair itself, created an initiating site for fatigue cracking to worsen.

ANALYSIS/SAFETY SIGNIFICANCE

The consequence of this condition was a violation of the Unit 1 Technical Specification (TS), Section 3.4.4, "Reactor Coolant System (RCS)," a reportable condition. The TS requires RCS leakage be limited to no pressure boundary leakage.

The safety significance of this condition is minimal. A 3/4-inch unisolable pipe leak from the reactor coolant pressure boundary from this condition is unlikely, given the drywell leakage observed prior to this repair, e.g., 0.6 GPM. Allowable reactor coolant system operational leakage limits are based on the predicted and experimentally observed behavior of pipe cracks. The probability is small that the imperfection or crack associated with such leakage would grow rapidly. The unidentified leakage flow limit of Technical Specifications allows time for corrective action before the reactor coolant pressure boundary could be significantly compromised. The five GPM limit is a small fraction of the calculated flow from a critical crack in the primary system piping. Crack behavior from experimental programs shows that leakage rates of hundreds of gallons per minute will precede crack instability. In summary, given the size of the leak, there were no actual consequences to the health and safety of the public.

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 internet 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.

CORRECTIVE ACTIONS

Completed - Repair of the cracked weld was performed prior to the restart of Unit 1.

Planned - The procedure for control of welding will be revised to include additional requirements for addressing repair of cracked welds on small bore reactor coolant pressure boundary piping, such that the cognizant weld engineer consults with the Non-Destructive Examination (NDE) supervision, and/or manager of programs engineering to:

• identify possible repair options with respect to removal of the weld metal defects (e.g., excavate unacceptable defects to sound metal, remove entire weld to base metal, etc.),

• assess risk for each identified repair option, (e.g., potential consequences if the method is ineffective),

• document selected repair option, and

• ensure the planner assigned to plan a weld repair work document is provided a specific repair option that documents their instruction.

PREVIOUS SIMILAR EVENTS

This event is similar to the following event, involving a weld crack repair and leakage at this same location.

• LER 50-387/2014-011-00: "Degraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by an Inadequate Weld," issued February 11, 2015.