05000286/LER-2013-004
| Event date: | 03-14-2013 |
|---|---|
| Report date: | 05-13-2013 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2862013004R00 - NRC Website | |
Note: The Energy Industry Identification System Codes are identified within the brackets {}.
DESCRIPTION OF EVENT
On March 14, 2013, during a scheduled refueling outage (RO) boric acid program walk down inspection, boron residue was identified on the fillet weld which attaches the E-11 in-core guide tube to the seal table. Since there was no visible leak path indicating that the leak had initiated at the high pressure mechanical seal connection located just above the fillet weld, the boron residue was cleaned and a surface examination [i.e., Liquid Penetrant Test (PT)] was performed on the weld. Although this surface examination did not identify any rejectable indications at the leak location, it did identify rounded indications in the fillet weld which could have been the cause of the leakage. No linear indications were identified on the weld or on the guide tube base metal above the toe of the weld. Because the PT results showed rounded indications in the weld and boron residue was present, the condition was judged to be an indication of a potential through wall defect therefore, a reactor coolant leak path. In August 2012 a maintenance walkdown identified dried boron residue at this location (CR-IP3-2012-02668). The boron was cleaned from the weld and a VT-1 (examination of imperfections) and a VT-2 (leakage) examination was performed. No active leakage or deformities were identified. The boron observed was limited, dry, and easily removed. No evidence of thru-wall defects were observed. Based on the results of the examinations Engineering concluded the area of interest was structurally sound with no reactor coolant pressure boundary (RCPB) leak. The seal table in-core detector guide tube is part of the RCPB. Technical Specification (TS) 3.4.13 (RCS Operational Leakage) does not allow any RCPB leakage. The condition was recorded in the Indian Point Energy Center (IPEC) Corrective Action Program (CAP) as Condition Report CR-IP3-2013-01556.
Inspections are performed during ROs on systems, components and piping inside containment that contain borated water and fall within the requirements of the Boric Acid Program. During the Unit 3 2013 spring RO, inspections identified locations where dry boron had accumulated. Each location was evaluated to determine if additional actions were required. As a result of these inspections, one location with boron deposits was judged to represent a through wall defect in RCPB. None of the remaining locations with boron deposits were attributed to RCPB leakage.
The E-11 in-core guide tube up to the seal table is approximately 1.0 inches in diameter up to the bottom of the seal table and then it reduces down to approximately 0.87 inches as it extends through the seal table up to the high pressure mechanical seal. The thimble tube, which extends into the core inside the reactor vessel, is located inside the guide tube. Reactor coolant fills the annulus between the thimble tube and the outer guide tube. The guide tube material is Type 304 Stainless Steel (SS). The fillet weld where the leak occurred is not a pressure boundary weld but rather a structural weld which attaches the outer guide tube to the seal table.
An extent of condition inspection consisted of visual (VT-2) inspections of the remainder of the guide tube to seal table fillet welds to determine if other signs of boron were also present. This visual inspection did not identify any additional issues at the remainder of these locations.
The Cause of Event The apparent cause of the through wall defect was OD initiated stress corrosion cracking (SSC) of the stainless steel (SS) guide tube base material under the fillet weld. This conclusion was based on finding no linear or other rejectable indications.
Engineering postulates that the rounded indications on the weld metal allowed contaminants from previous mechanical joint leaks to contact the guide tube base material under the weld resulting in SCC of the tube base material. Type 304 SS does not corrode when exposed to the RCS fluid. However, it is susceptible to stress corrosion cracking when contaminated with chlorides and/or fluorides. OD initiated SCC of SS caused by contaminants has been observed throughout the industry and is the most plausible defect propagating mechanism. ID initiated cracking of SS in PWR water chemistry environment is a lower probability mechanism given the low oxygen content in reactor coolant.
Corrective Actions
The following corrective actions have been performed under the Corrective Action Program (CAP) to address the cause of this event.
- A VT-2 visual examination of the remaining seal table penetrations was performed to verify that no additional defects existed. This examination confirmed that no additional defects existed.
- The defective guide tube was removed from service by cutting the tube below the leaking area and installing a welded plug to form the new reactor coolant pressure boundary. A subsequent inspection of the fillet weld and a pressure test confirmed the integrity of the guide tube and the new plug.
Event Analysis
The event is reportable under 10CFR50.73(a)(2)(i)(B) The licensee shall report any operation or condition which was prohibited by the plant's Technical Specification (TS). This event meets the reporting criteria because the Limiting Condition for Operation (LCO) for TS 3.4.13 allows no RCPB leakage and based on surface examinations and boron deposits, it was concluded the condition existed during past plant operation and the TS actions not taken.
Past Similar Events
A review was performed of the past three years of Licensee Event Reports (LERs) for events reporting a TS violation due to a through wall defect in the RCPB. No unit 3 previous LERs: LER-2012-003 and LER-2010-004.
455A Spray Inlet Stop valve 4152 contained a defect on the top of the horizontal leak- off pipe in the base metal approximately one inch from where the pipe connects to the valve bonnet. Defect #2) The socket weld of a 3/8 inch diameter tubing "tee" fitting down stream of valve 4138 contained a defect. The apparent cause of defect 1 (leak-off pipe) was stress corrosion cracking due to surface contamination. The apparent cause of defect 2 was poor quality weld due to insufficient weld reinforcement in part of the weld. Although the cause for defect #1 was similar the corrective action was to eliminate the capped leak-off pipes on RCS valves which would not have prevented the event reported in this LER.
valve 256B on the 22 Reactor Coolant Pump seal bypass line. The leak was discovered during a Refueling Outage inspection under the Boric Acid Program. The defect was a through wall indication as a result of a minor weld defect from the time of construction. This event is different as it was a result of missing weld material due to poor workmanship.
Safety Significance
This event had no significant effect on the health and safety of the public. There were no actual safety consequences from the event because there were no significant failures in the RCPB. Periodic inspections identify leaks when they are small so that repairs can be performed to prevent RCPB degradation. TS 3.4.13 has Surveillance Requirement 3.4.13.1 to verify RCS leakage is within limits by performance of RCS inventory balance every 72 hours. This surveillance ensures the integrity of the RCPB is maintained and provides a trend of leakage early before significant degradation. An early warning of RCPB leakage or unidentified leakage is provided by the systems that monitor containment atmosphere radioactivity and operation of the containment sump. Failure of a RCPB would be a Loss of Coolant Accident (LOCA). A LOCA is analyzed in UFSAR Section 14.3. A minor pipe break (small break) is defined as a rupture of the RCPB with a total cross-sectional area less than 1.0 square foot in which the normally operating charging system flow is not sufficient to sustain pressurizer level and pressure. The results of analysis in UFSAR Section 14.3.3.4 concluded the limiting break was a 3 inch cold leg break. The results of the analysis demonstrated that for a small break LOCA, the Emergency Core Cooling System will meet the acceptance criteria contained in 10CFR50.46. The LOCA analysis of Section 14.3 are bounding for the components reported in this LER.