05000287/LER-2003-001
| Oconee Nuclear Station, Unit 3 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 2872003001R00 - NRC Website | |
EVALUATION:
BACKGROUND
There are sixty-nine (69) Control Rod Drive (CRD) Mechanism (CRDM) [EIIS:AA] nozzles [El IS:NZL] that penetrate the Reactor Vessel Head (RVH) [EllS:RCT]. The CRDM nozzles are approximately 5-feet long and are welded to the RVH at various radial locations from the centerline of the RVH. The nozzles are constructed from 4-inch outside diameter (OD) Alloy 600 material.
The lower end of the nozzle extends about 6-inches below the inside of the RVH.
The Alloy 600 used in the fabrication of CRDM nozzles was procured in accordance with the requirements of Specification SB-167, Section II to the 1965 Edition including Addenda through summer 1967 of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code. The product form is tubing and the material manufacturer for the Oconee Nuclear Station (ONS) Unit 2 CRDM nozzles was the Babcock and Wilcox (B&W) Tubular Products Division.
Each nozzle was machined to final dimensions to assure a match between the RVH bore and the OD of each nozzle. The nozzles were shrunk fit by cooling to at least minus 140 degrees F., inserted into the closure head penetration and then allowed to warm to room temperature (70 degrees F minimum). The CRDM nozzles were tack welded and then permanently welded to the closure head using 182-weld metal. The manual shielded metal arc welding process was used for both the tack weld and the J-groove weld. During weld buildup, the weld was ground; and dye penetrant test (PT) inspected at each 9/32 inch of the weld. The final weld surface was ground and PT inspected.
The weld prep for installation of each nozzle in the RVH was accomplished by machining and buttering the J-groove with 182-weld metal. The RVH was subsequently stress relieved prior to the final installation of the nozzles.
EVENT DESCRIPTION
Oconee Nuclear Station Unit 3 (ONS-3) entered its scheduled end-of-cycle 20 refueling outage on April 26, 2003. On May 2, 2003, a visual inspection of the bare reactor vessel head (RVH) was performed, while bolted to the vessel, in order to determine if any of the sixty-nine (69) Control Rod Drive (CRD) Mechanism (CRDM) nozzle penetrations had developed a reactor coolant leak during the prior operating cycle. This inspection was performed looking through the nine access ports in the service structure support skirt on the RVH.
Results of the visual inspection revealed two (2) CRDM nozzles that were suspected of leakage. Of these, CRDM No. 4 was observed to contain a very thin white coating on the nozzle and CRDM No. 7 appeared to have a small accumulation of boron on the head adjacent to the annulus region. In addition, approximately 6 to 8 CRDMs could not be visually inspected as they were masked by deposits from a Component Cooling (CC) system [EllS:CC] leak above the RVH.
Subsequent evaluation of the prior refueling outage RVH inspection videotape showed that the CRDM No. 7 deposits were not associated with a new leak but rather were remnants from a prior refueling outage leak and repair campaign where the boron residue had not been completely removed from the RVH during the wash down process. The CRDM No. 4 boron deposit appeared fresher, exhibited characteristics similar to prior RVH leaks and as such, was conservatively identified as a leak. The apparent root cause of the nozzle leak is primary water stress corrosion cracking.
On May 2, 2003, after confirming that during power operations the Reactor Coolant System [EDS: RCS] pressure boundary had been degraded, an 8-hour notification (No. 39821) was made at 1957 hours0.0227 days <br />0.544 hours <br />0.00324 weeks <br />7.446385e-4 months <br /> (Eastern Time) in accordance with 10 CFR 50.72(b)(3)(ii)(A) reporting requirements.
ReportabilitV Technical Specification Limiting Condition for Operation 3.4.13(a) limits RCS operational leakage to "No pressure boundary leakage" while in MODES 1 through 4. This event also represents a degradation of one of the plant's principal safety barriers. Consequently, this event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(ii)(A) reporting requirements.
No operator intervention was required as a result of this event. Prior to the discovery of this event, Unit 3 was in cold shutdown (Mode 5) at 0 percent power and Units 1 and 2 were in Mode 1 operating at approximately 100 percent power.
ROOT CAUSE
Based on prior RVH evaluations described in previous, similar reported events (see below), the apparent root cause of the leaking Alloy 600 CRDM nozzle was Primary Water Stress Corrosion Cracking (PWSCC).
CORRECTIVE ACTIONS
The current RVH will be retired and replaced with a new RVH prior to ONS-3 restart.
SAFETY ANALYSIS
There were no actual safety consequences as a result of this event. The leakage of primary reactor coolant through the CRDM nozzle was so small that it was detectable only by the extremely small accumulation of boric acid crystals observed on the RVH. The total leakage from the CRDM nozzle did not exceed Technical Specification limits for unidentified RCS inventory loss. At no time during the operating cycle did the reactor building or area radiation alarms actuate as a result of this event.
Neither the small amount of boric acid crystal deposits observed around the nozzle nor the deposits from the Component Cooling system leak caused visible corrosion damage to the RVH.
NRC.FORI4 366A U.S. NUCLEAR REGULATORY COMMISSION (14301) FACILITY NAME (1) DOCKET (2) Since the current RVH will be retired from service and encapsulated for storage in the newly constructed steam generator / reactor head retirement facility, non-destructive examination (NDE) of the leaking nozzle was not performed primarily for personnel safety reasons and to minimize radiation exposure to workers in accordance with ALARA principles. The CRDM nozzle repairs made during the previous repair campaigns are shown on Figure 1.
ADDITIONAL INFORMATION
This event did not include a Safety System Functional Failure nor involve a personnel error. There were no releases of radioactive materials, radiation exposures in excess of limits or personnel injuries associated with this event. This event is considered reportable under the Equipment Performance and Information Exchange (EPIX) program. Energy Industry Identification System (EIIS) codes are identified in the text as [EIIS:XX].
SIMILAR EVENTS
Over the last two and one half years, similar event LERs have been submitted for all three Oconee units beginning with Unit 1 in December 2000 (LER 269/2000-06) and the last for Unit 2 in December 2002 (LER 270/2002-02). To date, three (3) Unit 1, two (2) Unit 2, and three (3) Unit 3 LERs, have been submitted to the NRC which have reported PWSSC of Alloy 600 CRDM and/or thermocouple nozzles (Unit 1 only). Prior to these LERS, there have been no other reportable events that involved PWSCC of Alloy 600 components or RVH penetration leaks.
Figure 1 — Oconee Unit 3 RVH Map CRDM Nozzles (69) � Thermocouple Nozzles (8) (Oconee Unit 1 Only)
- Nozzles 3, 7, 11, 23, 28, 34, 50, 56, and 63 were repaired (maintenance outage) 0 Nozzles 2, 10, 26, 31, 39, 49 and 51 were repaired (EOC 19 refueling outage)
- Nozzle 4 was not repaired since the RVH is being retired (EOC 20 refueling outage) 5