05000395/LER-2012-003
| Virgil C. Summer Nuclear Station Unit 1 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 3952012003R00 - NRC Website | |
ABSTRACT:
- On October 23, 2012 V. C. Summer Station Unit 1 (VCSNS) identified its first of four reactor vessel head penetrations that did not meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1. The station was in a refueling outage (RF20) and commenced volumetric testing of the reactor vessel head penetrations as a result of recent industry experience on other cold head plants. Even though the volumetric and visual inspections determined that four penetrations were not through wall, this event is reportable because of a material defect in the primary coolant system which cannot be found acceptable under ASME Section XI. Therefore, these inspection results are reportable pursuant to 10CFR50.73(a)(2) (ii)(A).
The flaws were subsequently repaired utilizing the embedded flaw repair process in accordance with NRC approved WCAP 15987, Revision 2-P-A. The apparent cause of the flaws is attributed to primary water stress corrosion cracking. Corrective actions included repairing the penetrations and revising the frequency of the Bare Metal Visual Exam and the Volumetric Exam to every refueling outage.
I. EVENT DESCRIPTION
During V. C. Summer Station Unit 1 (VCSNS) refueling outage, the station identified four reactor vessel head penetrations that did not meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1. The station was in a refueling outage (RF20) and commenced volumetric testing of the reactor vessel head penetrations as a result of recent industry experience on other cold head plants with similar material. Other plants with similar conditions occurring within the past 2 years include Byron 1, Byron 2, Braidwood 1 and Shearon Harris. These plants have head temperatures similar to VCSNS and share the same Alloy 600 tubular product manufacture. Shearon Harris is the "sister" vessel to VCSNS having both heats M6369 and M6370 which were found to have had Primary Water Stress Corrosion Cracking (PWSCC) during their Spring 2012 outage.
VCSNS inspected 65 penetrations and 1 vent line during RF-20 in accordance with ASME Boiler & Pressure Vessel (BPV) Code Case N-729-1. There have been no previous repairs to the reactor vessel head penetrations and/or j-groove welds. A visual examination was performed on the top of the head, which indicated no pressure boundary leakage. The examination of the reactor head was performed in accordance with the Electric Power Research Institute (EPRI) Performance Demonstration Initiative (PDI) program with qualified personnel, procedures and equipment provided by Wesdyne. The four penetrations did not have any indications that were determined to be through wall as indicated from volumetric and visual inspections. These four (4) penetrations are identified as 19, 31, 37 and 52, and all have the same Heat Number M6370.
II. EVENT ANALYSIS
The penetration indications experienced at VCSNS were found to be a result of Primary Water Stress Corrosion Cracking (PWSCC). Premature PWSCC initiation has been attributed to the results of residual strain in the material induced by cold work from rotary straightening during manufacture of the Babcock and Wilcox (B&W) Alloy 600 tubular products.
For VCSNS, the ultrasonic test length measurement indicated that the flaws existed within the tube material located inside the pressure boundary and extended up to the toe of the j-groove weld. Due to length measurement uncertainty, it was not certain whether these indications extend above the toe of the j-groove weld. The four CRDM nozzle indications that were suspect had a penetrant test performed on each indication. The indications in all four CRDM nozzles were confirmed to be surface connected. In each case, the penetrant test results confirmed the indications were in the base metal of the nozzle and not in the surface of the j-groove weld. The top of the head was visually examined in accordance with ASME Code Case N-729-1 with no indications found.
III. SAFETY SIGNIFICANCE
This condition had no actual safety consequences that were determined to impact the plant or public safety. The penetration flaws were identified prior to through wall leakage occurring and therefore were identified in a timely manner and immediately repaired. If the flaw remained undetected, it could have potentially propagated through the Alloy 600 weld material over time and formed a leak path through the reactor coolant pressure boundary.
IV. CORRECTIVE ACTIONS
The station obtained the NRC's permission [ML12325A432] to utilize WCAP 15987, Revision 2-P-A "Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations" [ML040290246 and ML0318402371.
The penetrations were repaired utilizing the embedded flaw repair process in accordance with WCAP 15987, Revision 2-P-A and NRC Temporary Verbal Relief authorization.
The station has changed the re-inspection interval to every refueling outage as required by 10CFR50.55a(g)(6)(ii)(D)(5).
V. PREVIOUS OCCURRENCES
No previous occurrences to this event have been identified.