05000313/LER-2016-004

LER-2016-004, Decay Heat Removal System Socket Weld Leak due to a Vibration-Induced Fatigue Crack

Arkansas Nuclear One, Unit 1
Event date:	09-30-2016
Report date:	11-29-2016
Reporting criterion:	10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Initial Reporting
ENS 52271	10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
3132016004R00 - NRC Website

LER 16-004-00 for Arkansas Nuclear One, Unit 1, Regarding Decay Heat Removal System Socket Weld Leak Due to a Vibration-Induced Fatigue Crack

ML16334A418
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/29/2016
From:	Anderson R L Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN111603 LER 16-004-00
Download: ML16334A418 (6)

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send 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 e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), 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 A. Plant Status At the time this condition was identified, ANO-1 was shut down and in a refueling outage with the core fully loaded.

No structures, systems, or components were out of service at the time of this event that contributed to this event.

B. Event Description

At approximately 20:47 on September 29, 2016, the ANO-1 Control Room received a report of a leak in the ANO-1 Reactor Building (RB). Investigation revealed a 0.125 gpm leak in a class 1 pipe from a socket weld [VTV] where the drain pipe connects to the main DHR system injection line. This leak was on a section of cross-connect piping shared by both trains of the DHR system.

Operations closed RB penetrations providing direct access from the RB atmosphere to the outside atmosphere in accordance with required actions of DHR Technical Specification (TS) 3.9.5.

To reduce the potential for propagation of the weld crack, Operations maintained DHR flow at less than 2000 gpm and monitored data from vibration probes affixed to the affected piping to maintain low vibration levels.

To remove the reliance on the DHR system, actions were taken to defuel the reactor vessel, which was accomplished at 12:00 on Thursday, October 6, 2016. The leak remained less than or equal to 0.125 gpm from the time of discovery until the piping was removed from service for repair.

C. Background — System Design The DHR /Low Pressure Injection (LPI) [BP] pumps have three major functions. The first is to remove decay heat from the fuel and Reactor Coolant System (RCS) once the plant is shut down and cooled down. To do this, the DHR pump takes suction from the RCS, discharges through a cooler which is cooled by Service Water, and injects the cooled water back into the RCS. A second function is to provide required net positive suction head from the RB Sump to the High Pressure Injection (HP1) pumps in the event of an intermediate size loss-of-coolant accident (LOCA). The third function is LPI of borated water from the Borated Water Storage Tank or RB Sump into the RCS in the event of a large-break LOCA or after cooling down and depressurizing sufficiently following an intermediate break LOCA.

D. Event Cause This condition was self-revealing. The direct cause of the leak was a fatigue crack caused by vibration of the drain.

ANO's corrective actions for previous DHR/LPI system drain line socket weld cracks did not either adequately address worst case operating conditions or eliminate the source of the vibration. In addition, the DHR operating procedure contained inadequate administrative barriers to minimize DHR system vibration. Furthermore, previous corrective actions did not ensure that Operations training programs were updated regarding the basis for limiting DHR system flow to minimize system vibration.

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send 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 e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), 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 50-000313 Extent of Condition The socket welds on other drains and vents in the ANO-1 DHR system were evaluated which resulted in cut out and replacement of other vents and drains. The drain line socket welds were replaced with enhanced welds. The tie-back supports for DH-1037 and DH-1450 were designed using as-measured vibration spectra to provide hardening against system vibration. The evaluation of other systems and components identified no additional corrective actions for vibration-induced damage.

E. Corrective Actions

The corrective actions taken and planned are:

• The cracked socket weld was repaired. The tie-back drain line support was redesigned and replaced. This hardens the system against vibration-induced fatigue cracking and component damage.

• The DHR system piping and supports inside the RB were inspected for damage. No other damage was identified. The stresses on the DHR system one-inch drain line isolation valves inside the RB were calculated using as-measured response spectra to ensure acceptable performance. The socket welds for several other drain valves were cut out and replaced using enhanced welds.

• Post-modification testing of the drain valve and support modifications confirmed that the corrective actions would result in acceptable vibration of the drain lines during high system flow rates.

• The DHR operating procedure is planned to be revised to provide specific administrative guidance to minimize DHR system vibration, consistent with results of post-modification testing.

F. Safety Significant Evaluation For the function of the DHR system, the effect of the one-inch drain line completely separating from the eight-inch Core Flood Tank injection line is estimated to be a leak of 98 gpm with the backpressure that would exist at 4,000 gpm of DHR system flow. When the leak was discovered, both DHR pumps were in service. The second DHR pump was secured and the remaining pump flow was throttled back to approximately 2000 gpm (2000 gpm is adequate for core cooling at all expected decay heat levels). The maximum flow that can be provided by a DHR/LPI pump is 3547 gpm per CALC-92-E-0077-08. Therefore, the potential loss of approximately 175 gpm through the leak path would not have prevented either DHR train from providing adequate core cooling.

For the function of piggyback operation (suction from RB Sump through LPI Pump to HPI pump) the HPI pump has a maximum flowrate of approximately 525 gpm. Considering the aforementioned LPI flow capability, complete separation of the one-inch drain line would not have prevented the HPI pump from performing its specified safety function.

In the LPI mode of operation following a trip from full power, the mechanism which caused the flow-induced vibration fatigue failure would not have occurred until RCS pressure had lowered to approximately 30 psig. This is because venturi cavitation does not occur until the pressure downstream of the venturis reaches approximately 30 psig. The extent-of-condition review identified three drains that were susceptible to this vibration-induced fatigue failure. The total loss of the three drains equates to an area of about 1.56 sq. inches, which is bounded by supporting LOCA calculations. A review of the associated pump curves indicated that there is more than enough margin to account for the losses through the drain lines assuming these two failed during a LOCA.

Reported lessons learned are incorporated into the licensing process and fed back to industry.

Send 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 e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), 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 Arkansas Nuclear One, Unit 1 50-000313 The significance of the leak is lessened by the fact that the leak is located inside the RB and any leaks are fully recoverable by aligning the suction of the DHR/LPI pump to the RB Sump.

Summary:

The actual consequences as stated in the problem statement were that both trains of DHR were declared inoperable. There were no other actual consequences to general safety of the public, nuclear safety, industrial safety, and radiological safety for this event.

The potential consequences to general safety of the public, nuclear safety, industrial safety, and radiological safety of this event are reduced flow to the core in the event of a large-break LOCA. However, adequate flow remains available to ensure the reactor can be maintained in a safe shutdown condition.

Based on the above, the significance of this condition is low.

G. Basis for Reportability 10 CFR 50.73(a)(2)(v)(B): Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat.

10 CFR 50.73(a)(2)(v)(D): Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.

At the time of discovery, both trains of the DHR system were in service. Based on the location of the leak both trains of DHR were declared inoperable. With both trains being declared inoperable the associated safety function could not be met. In accordance with the NUREG-1022 guidelines, at the time of discovery ANO-1 made the 8-hour notification under EN 52271 and this LER is required.

H. Additional Information

10 CFR 50.73(b)(5) states that this report shall contain reference to "any previous similar events at the same plant that are known to the licensee." NU REG 1022 reporting guidance states that term "previous occurrences" should include previous events or conditions that involved the same underlying concern or reason as this event, such as the same root cause, failure, or sequence of events.

A review of the ANO corrective action program and LERs for the previous three years was performed. No relevant LER were similar to the failure mode.