05000483/LER-2013-006
| Callaway Plant Unit 1 | |
| Event date: | 05-08-2013 |
|---|---|
| Report date: | 07-03-2013 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 4832013006R00 - NRC Website | |
1. DESCRIPTION OF STRUCTURE(S), SYSTEM(S) AND COMPONENT(S):
The Accumulator Safety Injection (EP) system [EIIS Code: BP] functions to deliver borated water from an accumulator tank [EIIS code ACC] installed on each of the four Reactor Coolant System (RCS) [EIIS Code: AB] cold legs during the post-LOCA injection phase in order to mitigate the consequences of a design basis accident (DBA). Each accumulator is connected to its respective RCS cold leg piping through a ten-inch pipe. Accumulator D is equipped with a 3/4-inch manual vent valve [EIIS code V] that is used to support maintenance activities for the accumulator and is connected to the six-inch injection lines from the Residual Heat Removal (RHR) system [EIIS Code:
BP] and high head safety injection pump [EIIS Code: BQ] injection line, which connect to the ten- inch accumulator discharge line. A 3/8-inch diameter orifice [EIIS code OR] is drilled into the pipe wall within the fitting at the vent piping connection and defines the boundary between the American Society of Mechanical Engineers (ASME) Code Class 1 RCS piping and the ASME Code Class 2 accumulator vent piping. The orifice ensures that flow through this line, in the event of a catastrophic guillotine-type break, is within the capability of a normal charging pump. The vent valve is normally closed and capped during normal operation. Valve EPV0109 is the 3/4-inch vent valve for accumulator tank TEPO1D.
2. INITIAL PLANT CONDITIONS:
On May 8, 2013, the plant was in a refueling outage, and at the time of the event, in Mode 6, Refueling. The core was reloaded and the refueling pool was flooded up greater than 23 feet above the reactor vessel flange. In this operating Mode, only one train of Residual Heat Removal (RHR) was required to be operable and in operation. The B train of RHR was in operation in the cooldown lineup.
3. EVENT DESCRIPTION:
On May 8, 2013 at approximately 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> CDT, water was observed dripping from the insulation on piping connected to Reactor Coolant System (RCS) loop 4. Further investigation determined it was near Accumulator Safety Injection vent valve EPV0109. Subsequent to the observation, scaffolding was installed and insulation around the valve and associated piping was removed to facilitate inspection.
At approximately 0509 hours0.00589 days <br />0.141 hours <br />8.416005e-4 weeks <br />1.936745e-4 months <br /> CDT on May 9, 2013, Engineering inspected the piping and determined there was a crack in the socket weld where 3/4-inch vent valve EPV0109 is connected to the B train injection piping that connects to the cold leg of RCS. The estimated leakage rate through the crack was 6 drops per minute. The B RHR train was taken out-of-service in response to the identified condition.
A repair was planned to remove the cracked section of pipe by removing the weld at the sockolet coupling, shortening the pipe, and reinstalling the weld using a 2-to-1 fillet weld leg configuration.
The repair was performed on May 10, 2013.
4. ASSESSMENT OF SAFETY CONSEQUENCES:
At the time the leak at valve EPV0109 was discovered, the plant was in Mode 6 with the refuel pool flooded to a level greater than 23 feet above the reactor vessel flange. The identified through wall crack allowed leakage from the primary coolant system. The boron concentration of the RCS was greater than 2450 ppm boron. Plant Technical Specifications required the boron concentration to be ? 2000 ppm boron.
The leak rate due to the cracked pipe was estimated to be six drops per minute. In the event of a complete pipe failure the maximum leak rate at this location would have been limited by a 3/8-inch flow restrictor that is located in the wall of the six-inch pipe at the EPV0109 branch connection.
This flow restrictor is designed to limit the leak rate, in case of a complete failure of the connection on the vent line, to a value below the capacity of a single charging pump at normal operating pressure and temperature. This design feature would allow the plant to achieve and maintain safe shutdown under all operating conditions even considering a complete failure of the welded connection.
The A RHR train was operable at the time that the B train was declared inoperable due to the leak at EPV0109. For the plant conditions that existed at that time, one operable train of RHR would have been sufficient to remove decay heat. The ability to control a potential release of radioactive material or mitigate the consequences of an accident was not impacted by this event.
Based on the above, this condition is considered to have very low safety significance.
5. REPORTING REQUIREMENTS:
The leak from the piping around valve EPV0109 was reported to the NRC via Event Notification (EN) 49014 on May 9, 2013.
This event described in this License Event Report (LER) is a condition that constituted degradation of a principal safety barrier and is considered reportable to the requirements of 10 CFR 50.73(a)(2)(ii)(A). Guidance provided in section 3.2.4 of NUREG 1022, Revision 2, states that conditions that represent "welding or material defects in the primary coolant system which cannot be found acceptable under ASME Section XI, IWB-3600, 'Analytical Evaluation of Flaws,' or ASME Section XI, Table IWB-3410-1, 'Acceptance Standards,' are reportable to this criterion.
6. CAUSE OF THE EVENT:
The most probable cause was that the configuration of the B RHR train results in vibration levels that can damage equipment. In this case, the system vibration levels were strong enough to induce cyclic fatigue at the socket weld upstream of valve EPV0109. Vibration testing was performed as part of the post-maintenance testing for the May 10, 2013 repair mentioned above. The testing identified that vibration levels exceeded OM Code part 3 screening criteria.
A contributing cause was that the design geometry for the branch connection (the connection of the vent line to the injection piping) allowed stress to concentrate at the socket weld, which contributes to fatigue failures.
7. CORRECTIVE ACTIONS:
The cracked section of pipe was removed by removing the weld at the sockolet coupling, the pipe was shortened, and a weld was reinstalled using a 2-to-1 fillet weld leg configuration, which was recommended for improved fatigue resistance. (Note: a 2-to-1 fillet weld configuration means the weld leg on the pipe side of the weld is twice as long as the leg on the fitting side.) This work was performed on May 10, 2013.
Vibration testing was performed as part of the post-maintenance testing for the repair mentioned above on May 10, 2013. The testing identified that vibration levels exceeded OM Code part 3 screening criteria. A plant corrective action document was initiated, and a plant modification was then developed to install a support brace to the branch connection at EPV0109. The modification was installed on May 20, 2013. Vibration levels at the branch connection were reduced to acceptable levels.
As an extent of condition/cause review during Refuel 19, previous piping vibration analyses were reviewed and screened against like-kind valve arrangements (vent/drain valve connections for the ECCS systems inside the Containment building). Four valves were identified where the previously measured vibration levels were 80 percent or higher of the allowable calculated vibration levels. Two of the identified valves were visually inspected with no leakage identified. Penetrant testing was performed on the third valve with satisfactory results. Vibration testing was performed on the fourth valve and was evaluated to be acceptable.
The above actions addressed the immediate issue of the leak. Long-term actions to address the B RHR train vibrations levels and the branch connection design geometry include the following:
Eliminate or reduce sources of vibration on B RHR train from the pump discharge to the RCS such that no branch connection exceeds 80 percent of ASME OM Part 3 calculated acceptance criteria at worst case operating conditions, if possible.
If vibration cannot be reasonably eliminated or reduced below 80 percent of ASME OM Part 3 calculated acceptance criteria as described above, then manage vibration by upgrading the socket welds or modifying vulnerable piping configurations.
Create or revise engineering guidance documents to provide guidance about robust designs or systems with high vibration in order to improve fatigue resistance.
8. PREVIOUS SIMILAR EVENTS:
In November 2003, Wolf Creek experienced a similar leak at the same location, EPV0109. The event is described in Wolf Creek LER 2003-004-00, Failure of Safety Injection Accumulator Vent Line.
On October 26, 2002, a leak was identified at a seal weld joint in the flanged connection to valve EJV0179, RHR TRN B ACC INJ LOOP 3 EJF00003 UPSTRM TEST CONN. The physical cause was determined to be a fatigue crack due to high stress concentration in combination with an inadequate joint design.
In February 2002, high vibration was identified in the piping associated with the motor driven auxiliary feedwater trains.
Callaway Plant Unit 1 05000483