05000387/LER-2012-007
| Susquehanna Steam Electric Station (Sses) Unit 1 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| Initial Reporting | |
| ENS 48036 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown, 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| 3872012007R00 - NRC Website | |
CONDITION PRIOR TO THE EVENT
Unit 1 — Mode 1, 100 percent Rated Thermal Power There were no inoperable structures, systems, or components that contributed to this event.
EVENT DESCRIPTION AND TIMELINE
A chronological timeline or sequence of events leading up to and immediately following the event follows:
On 06/15/2012, Operations personnel identified an increasing trend on the Unit 1 Containment Radiation Monitor (CRM) particulate channels. Over the next 3 days, Unit 1 unidentified drywell leakage continued to slowly rise. An Adverse Condition Monitoring Plan was developed for monitoring the leakage rate, and an Operational Decision Making (ODM) document was developed to drive conservative actions.
On 06/19/2012 a controlled Unit 1 shutdown was initiated prior to reaching ODM limits because the unidentified drywell leakage increased at a higher than expected rate. A low power containment entry was performed to identify the source of the leak which was a through wall leak in the "1A" reactor recirculation pump suction line decontamination flange.
On 06/19/2012 at 1720 hours0.0199 days <br />0.478 hours <br />0.00284 weeks <br />6.5446e-4 months <br />, Event Notification48036 was made to the NRC in accordance with 10 CFR 50.72(b)(2)(i), "the initiation of any nuclear plant shutdown required by the plant's Technical Specifications" and 10 CFR 50.72(b)(3)(ii)(A) "any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded." As such, this Licensee Event Report is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(A) and 10 CFR 50.73(a)(2)(ii)(A).
CAUSE OF THE EVENT
The direct cause of the through wall leak was a crack in the heat affected weld area of the connection (weldolet) joining the 4 inch diameter chemical decontamination pipe to the "1A" reactor recirculation pump suction line. Based on metallurgical examinations, the outside diameter (OD) initiated crack was due to cyclic fatigue.
The preliminary cause of the through wall crack is cyclic fatigue due to vibration caused by operating the reactor recirculation pumps near the decontamination flange assembly natural frequency. The normal operating speed range for the reactor recirculation pumps correlates to a 5x vane pass frequency of 122 Hz to 135 Hz. The configuration of the reactor recirculation pump suction line decontamination flanges is such that its natural frequency is 135.8 Hz.
Having the 4 inch pipe natural frequency near the pump 5x frequency results in vibration accelerations that are greatly magnified, relatively large bending moments, and stresses at the weldolet connection that may exceed the endurance limit of the material which can lead to fatigue failure.
An OD initiated crack, approximately 3 inches long, was identified just below the weldolet to 4 inch pipe transition weld. The crack was located in the heat affected zone of the connection (weldolet) joining the 4 inch diameter chemical decontamination pipe to the "1A" reactor recirculation pump suction line. In addition, inside diameter (ID) initiated cracks were identified in the weldolet.
Metallurgical lab analysis results were that the crack was caused by fatigue. The crack OD length was approximately 3 1/8 inches and the ID length was 2 5/8 inches. This is the length of the through wall portion of the crack which originated from the pipe OD. The crack was not caused by intergranular stress corrosion cracking (IGSCC), however indications of IGSCC were found in the weld unrelated to the crack.
Upon completion of the root cause evaluation for this event, the root cause will be provided in a supplement to this LER.
ANALYSIS/SAFETY SIGNIFICANCE
Actual Consequences:
The actual safety consequences of this event are minimal. An orderly shutdown of the Unit 1 reactor commenced when unidentified leakage reached approximately 1.1 gpm. The shutdown was performed well before the unidentified drywell leak rate reached the Technical Specification (TS) 3.4.4 limit of 5 gpm.
Potential Consequences:
The potential safety consequence of this event is that, if not corrected, this leak could have progressed to the point of reaching critical crack size. Reaching the critical crack size would have resulted in a PRA initiating event such as a scram or mid sized loss of coolant accident (LOCA). This could have resulted in an emergency plan entry for excessive drywell leakage and may have resulted in reactor coolant system loss through a 4 inch pipe. However, this potential consequence is unlikely to have occurred. There is significant margin between the 5 gpm maximum unidentified drywell leakage allowed by TS 3.4.4 and the leakage expected prior to reaching the criticar crack size.
Upon completion of the root cause evaluation, the actual safety significance and any potential consequences will be included in a supplement to this LER.
CORRECTIVE ACTIONS TAKEN
An immediate corrective action taken was to modify the chemical decontamination assembly connected to the reactor recirculation piping. This consisted of a redesigned 4 inch diameter chemical decontamination flange connection on both the "1A" and "1 B" reactor recirculation pump suction lines. The length of the 4 inch diameter pipe was reduced from 6 inches to approximately 3.5 inches to produce a new configuration that is not susceptible to the cyclic fatigue caused by reactor recirculation pump vane passing frequency that is very close to the natural frequency of the assembly. This modification also eliminated the IGSCC that were identified during metallurgical examinations.
Extent of condition inspections were performed to provide assurance that similar reactor recirculation and reactor water cleanup system piping has not been similarly affected. The inspections were completed, no vibration related issues were identified, and no additional fatigue flaws were identified in the expanded scope population.
An in-service leak test was conducted in accordance with the requirements of ASME Section XI at a test pressure of 1035 psig to ensure pressure boundary integrity of the modified decontamination flange connections.
As left impact resonance tests were performed on the "1A" and "1 B" reactor recirculation suction pipe decontamination flanges utilizing a test acceptance criteria for natural frequency greater than or equal to 150 Hz.
As a compensatory action for SSES Unit 2, the Unit 2 reactor recirculation pump speed is limited to 1600 rpm to provide margin to the resonant frequency of the reactor recirculation decontamination connection assemblies.
CORRECTIVE ACTIONS PLANNED
Completion of the critical flaw size analysis to determine the continued growth crack size at which application of design basis loads would have resulted in failure.
Additional corrective actions determined during completion of the root cause evaluation will be included in a supplement to this LER.
PREVIOUS SIMILAR EVENTS
There are no Susquehanna LER's related to TS required shutdown due to unidentified drywell leakage or reactor recirculation system leakage.
Industry Operating Experience (0E20348) for a similar event at Hope Creek LER 354/2005-002-00 was reviewed to ensure an understanding of the causes and actions taken. The investigation, inspection, and repair activities performed by SSES are consistent with those performed by Hope Creek.