05000265/LER-2014-001
| Quad Cities Nuclear Power Station Unit 2 | |
| Event date: | 03-31-2014 |
|---|---|
| Report date: | 05-30-2014 |
| 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 49977 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown, 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| 2652014001R00 - NRC Website | |
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 Internet 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 cfisplay 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.
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
Control Rod Drive Hydraulic Control Unit Scram Inlet Isolation Valve Pressure Boundary Leak Discovered During Inservice Inspection Program VT-2 Examination
A. CONDITION PRIOR TO EVENT
Unit: 2 Reactor Mode: 1 Event Date: March 31, 2014 Event Time: 1302 hours0.0151 days <br />0.362 hours <br />0.00215 weeks <br />4.95411e-4 months <br /> Mode Name: Power Operation Power Level: 86%
B. DESCRIPTION OF EVENT
On March 31, 2014, at 1302 hours0.0151 days <br />0.362 hours <br />0.00215 weeks <br />4.95411e-4 months <br />, an Inservice Inspection (ISI) Program VT-2 examination of the Unit 2 Control Rod Drive (CRD) [AA] Hydraulic Control Unit (HCU) [HCU] ASME Class 2 piping and components was being performed.
Engineering personnel performing the inspection discovered an apparent through-wall leak on the 2-0305-101-18-27 CRD HCU Scram Insert Isolation Valve [ISV] (the "101" valve). The leakage was noted to be approximately two drops per minute (dpm), and appeared to be seeping from the valve body, on the opposite side of the valve bonnet at approximately 1/2 inch above the center of the valve seat. This valve is subjected to full reactor [AD] pressure during normal service and during this inspection. This valve is the isolation valve to the reactor vessel CRD drive housing, and since it is the first isolation boundary off of the reactor vessel, it therefore cannot be isolated from the reactor coolant system to allow repairs. Operations declared the 101 valve inoperable and entered Technical Specifications (TS) LCO 3.4.4 Condition C.
An attempt was made to isolate the leak by shutting the 101 valve, but this was not successful.
At 1411 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.368855e-4 months <br />, Operators initiated an emergent load drop (between 30 and 40 MWe) to insert control rod E-7 in an attempt to isolate the leak on the 2-0305-101 valve.
At 1416 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.38788e-4 months <br />, Operations declared control rod E-7 (HCU 18-27) inoperable due to the through wall leak on the 2- 0305-101 valve and TS LCO 3.1.3 Condition C was entered.
At 1444 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.49442e-4 months <br />, control rod E-7 (HCU 18-27) was inserted to position 00 and isolated and disarmed.
At 1615 hours0.0187 days <br />0.449 hours <br />0.00267 weeks <br />6.145075e-4 months <br />, Operations started Unit 2 shutdown to effect repairs on the 2-0305-101 valve (Unit 2 Forced Outage, Q2F66 commences).
At 1757 hours0.0203 days <br />0.488 hours <br />0.00291 weeks <br />6.685385e-4 months <br />, ENS#49977 was made to the NRC under 10 50.72(b)(3)(ii)(A) since the defect was associated with the reactor coolant system pressure boundary. The notification was also made in accordance with 10 CFR 50.72(b)(2)(i) given the initiation of a plant shutdown required by the plant TS.
On April 1, 2014, the 2-0305-101 valve was removed from the system and shipped to PowerLabs for analysis.
Subsequently, PowerLabs determined that the through wall leak that developed was the direct result of an inherent manufacturing defect that eventually propagated to the surface following years of pressure and temperature cycles that the system normally experiences. The defect was described as an angled, non-branching flaw that followed the grain deformation lines and was most likely a partially sealed seam or lap created during the manufacturing process.
There was no evidence that a long-term corrosion mechanism, such as intergranular stress corrosion cracking (IGSCC) had contributed to the propagation of the identified defect.
Unit 2 commenced startup on April 1, 2014, following successful replacement of the valve. Inspection validated the integrity of the repair, which is a Class 2 pressure boundary.
Due to the impact on the reactor coolant pressure boundary, this report is submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(ii)(A), which requires the reporting of any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. Since a plant shutdown was completed as required by the plant TS, this report is also submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(A), which requires the reporting of the completion of any nuclear plant shutdown required by the plant's TS.
C. CAUSE OF EVENT
Exelon PowerLabs performed a metallurgical analysis on the removed valve. The analysis determined that the leakage was caused by an angled, non-branching manufacturing defect that followed the grain deformation lines that had opened to the surface due to service pressure/temperature cycles. The defect was most likely a partially sealed seam or lap that occurred during the forging process. There was no evidence that stress corrosion cracking contributed to defect propagation. The valve was installed during original plant construction and was in service over 42 years prior to its failure. No contributing factors were identified that required additional corrective action.
D. SAFETY ANALYSIS
System Design The purpose of the reactor pressure vessel (RPV) and its appurtenances, such as this section of the non-isolable ASME Class 2 hydraulic riser, is to retain the reactor core coolant-moderator within the RPV during all modes of plant operation.
The 101 valve is a manual gate isolation valve positioned on each CRD hydraulic insert riser. The 101 valve serves as the isolation valve to the HCU (isolates the scram inlet from the CRD). Each of the 177 CRD HCUs (per unit) has a 101 valve, as well as a 102 valve which isolates the withdrawal riser (isolates the scram outlet from the CRD). Both the 101 and 102 are the Class 2 pressure boundary for isolation from the reactor coolant system. Each HCU contains several additional isolation valves from the same manufacturer which are not part of the reactor coolant system pressure boundary.
On a control rod insertion, drive water flows up this riser to the under piston area of the Control Rod Drive Mechanism (CRDM) at a pressure high enough to drive the CRDM against reactor pressure (typically 1260 psig at full reactor power). On a rod withdrawal, exhaust water at reactor pressure flows down this riser from the under piston area of the CRDM. During periods of no rod motion, a small amount of cooling water continuously flows up this riser at just over reactor pressure (typically 1015 psig at full reactor power). On a reactor scram, the scram inlet valve opens a flow path from the accumulator to the under piston area of the CRDM via this header.
Safety Impact The safety significance of this event was minimal since the leakage rate was very small and full scram capability was maintained by the control rod. The leak was seeping at a rate of two dpm. The water inside the valve was pressurized at approximately 1000 psig; there was no spray or steam plume emitted. The CRD system including the 101 valve is visually inspected once per quarter. The last inspection was performed on February 28, 2014, with no leakage identified. After the leak was identified, Unit 2 was shutdown with no complications.
This associated failure mechanism that caused the valve leak was determined to be highly localized and slow acting.
This valve was installed during initial plant construction and the manufacturing defect did not propagate to the surface until after 42 years of service (approximately 30 Effective Full Power Years). The defect was observed to not result in a crack that was likely to propagate to the weld. Therefore, this defect was unlikely to result is a sudden large leak or catastrophic failure of the valve. The quarterly visual inspection performed on these valves is an adequate frequency to identify a valve body leak in sufficient time before any large leak can develop. Furthermore, each valve is VT-2 inspected for leakage during the periodic ISI pressure test, which occurs three times during each ten year interval.
Had the valve body catastrophically failed, such that there was a 1-inch diameter opening in the hydraulic riser connected to the RPV, the consequences would have been minimal. The 1-inch diameter line break (0.01 sq ft in area) would be bounded by the reactor coolant system line breaks of up to 0.12 sq ft in area, as discussed in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR), which provides that the High Pressure Coolant Injection (HPCI) system could supply sufficient coolant to depressurize the vessel and cool the core for line breaks of up to 0.12 sq ft in area (small break LOCA). If needed for emergency insertion, the control rod is capable of being scrammed using reactor pressure only (no external driving force on the under piston side of the control rod drive through the insert riser is needed).
The as-found leak at the insert riser isolation valve had no effect on the ability of the control rod to be inserted via normal means or scram insertion during an emergency. No changes were observed in CRD HCU system pressure or flow that indicated the two dpm leak had any impact on system performance.
Risk Insights Prior to and during this event, the capability of the CRD system, including the ability to insert this control rod (E-7) was not lost. This condition did not create any actual plant or safety consequences since the Unit was not in an accident or transient condition during this period of time or prior operating cycle. The unit was taken offline and placed into a shutdown condition without incident to effect repairs on this valve and for compliance with TS.
There is no makeup rate required to mitigate a RCS leakage of two dpm. The CRD HCU insert line at the isolation valve is a 1-inch pipe/fitting. Even with a complete failure of the valve body, this would only result in a Small LOCA (SLOCA). The break diameter of 1 inches would be well within the capability of HPCI or one Residual Heat Removal (RHR) pump. The 2 dpm leak is negligible compared to any SLOCA.
Based on the above, the change in risk due to a two dpm leak is negligible. Therefore, considering the impact of this condition on the Plant Probabilistic Risk Assessment (PRA), the change in Core Damage Frequency (CDF) due to the observed leakage will be less than 1.0E-06/yr. In conclusion, the overall safety significance and impact on risk of this event was minimal.
E. CORRECTIVE ACTIONS
Immediate:
1. Unit 2 was shutdown and the HCU insert manual isolation 101 valve was cut out and replaced, then inspected by qualified personnel prior to Unit startup.
2. All manual isolation valves associated with the RCS pressure boundary on Unit 2 were VT-2 inspected. No additional leaks were found. For Unit 1, the last VT-2 inspection was performed on these valves in 2011.
Operations walkdowns are performed daily on each Unit which include the CRD HCUs, and System Manager's walkdowns are performed quarterly on each Unit. No leaks have been identified on Unit 1.
Follow-up:
1. A Root Cause Investigation was initiated to determine the valve failure mechanism and corrective actions.
2. The failed valve was sent to Exelon PowerLabs for analysis. The analysis revealed a latent manufacturing defect that eventually caused the valve to fail and leak after over 42 years of service.
3. Related 101 valve classification and inspection procedures are being reviewed for potential enhancements.
F. PREVIOUS OCCURRENCES
The station events database, LERs, and INPO Consolidated Event System ICES (EPIX) were reviewed for similar events at Quad Cities Nuclear Power Station. This event was a CRD HCU isolation valve through-wall leak at the reactor coolant system pressure boundary. There were no other previous similar occurrences identified at Quad Cities Nuclear Power Station that were associated with this type of failure.
G. COMPONENT FAILURE DATA
Failed Equipment: 1 inch manual gate valve Component Manufacturer: Hancock Mfg. Co., Inc.
Component Model Number: 1-950W-3-XOSH10 Component Part Number: 131C9096 Component Type: 1 inch gate valve, F316 stainless steel This event has been reported to ICES as Report No. 310466.