05000302/LER-2003-003
| Docket Number | |
| 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 |
| Initial Reporting | |
| ENS 40222 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| 3022003003R00 - NRC Website | |
pressurizer spray was initiated. During Refueling Outage 12 in October 2001, the pressurizer spray valve [AB, V] was repacked and the insulated pressurizer spray line [AB, PZR] received a focused walk down, in addition to the routine boric acid and Reactor Coolant System (RCS) leakage walk downs. After returning to power operations, the RB airborne activity still increased when spray was initiated. A team was formed to address the multiple concerns raised by this phenomenon. The team's activities are documented in the CR-3 Corrective Action Program under Nuclear Condition Report (NCR) 69214. Data gathered on-line, walk downs during forced outages, and external operating experiences led to identifying the pressurizer steam space as the most likely source of the radionuclides escaping into the RB. Therefore, the most likely leak sites were nozzles, flanges or piping directly connected to the pressurizer steam space.
Focused inspection teams were formed to inspect selected locations on the pressurizer and connected piping during Refueling Outage 13. On October 4, 2003, initial inspections were performed in MODE 3 because the higher plant pressure would facilitate leak detection. Ultrasonic guns were used to enhance the ability to acoustically detect a leak. During the walk downs, ultrasonic noise was detected at the joint between the RC-1-LT3 upper level instrument tap nozzle and the pressurizer shell. Similar noise was detected in the vicinity of the upper level instrument tap nozzle for RC-1-LT1. After plant conditions allowed insulation removal, a visual inspection was performed.
EVENT DESCRIPTION
At 1032, on October 4, 2003, Progress Energy Florida, Inc., (PEF's) Crystal River Unit 3 (CR-3) was in MODE 4 (HOT SHUTDOWN) at zero percent RATED THERMAL POWER. While performing a visual inspection of pressurizer upper level instrument tap nozzles RC-1-LT1, RC-1-LT2 and RC-1-LT3 [AB, NZL], evidence of a very small RCS pressure boundary leak was found at each nozzle. The leakage evidence consisted of light boric acid deposits and stains on the pressurizer carbon steel shell for the RC-1-LT1 and RC-1-LT3 nozzles. Stains, but no boric acid deposits, were found on the pressurizer carbon steel shell for the RC-1-LT2 nozzle. No leakage evidence was found on any of the similar nozzles: the lower level taps, resistance temperature detector (RTD) thermowell, water sample tap or vent nozzle. The inspection was performed to determine the source of increased RB airborne activity noted when pressurizer spray was initiated.
Improved Technical Specification (ITS) 3.4.12 states that RCS operational leakage shall be limited to no pressure boundary leakage in MODES 1, 2, 3, or 4. If pressure boundary leakage exists, the plant shall be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (Action B.1) and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Action B.2).
At 1741 on October 4, 2003, CR-3 entered MODE 5.
serves the following functions: (a) establishes and maintains RCS pressure; (b) provides a surge volume for limiting pressure excursions; and, (c) provides a water reserve to accommodate reactor coolant temperature/density changes during operation. The primary safety function provided by the pressurizer is to maintain RCS pressure boundary integrity.
Industry operating experience with pressurizer nozzle cracks dates to the 1980's. Experience shows that periodic visual inspection detects the cracks before they grow large enough to challenge the structural stability of the nozzle. In all cases, when metallurgical failure analyses were performed, the cracking was axial in direction. There is no evidence that a pressurizer instrument penetration has ever developed a significant circumferential flaw. The RCS unidentified leakages never approached the 1 gallon per minute (gpm) Technical Specification shutdown limit.
Since failure analyses have indicated that cracking is in the axial direction, there is minimal risk of a circumferential break of the nozzle. Relatively low axial stresses, in comparison with radial stresses, in these small nozzles significantly reduce the possibility of developing circumferential cracks, even after leakage has developed.
In the case at CR-3, the RCS unidentified leakage rate was very small and never approached the 1 gpm Technical Specification limit. There was no appreciable accumulation of boric acid or wastage of the carbon steel shell, which was confirmed by visual inspection of the pressurizer shell around the three leaking nozzle penetrations and ultrasonic testing of the base material. This is consistent with Operating Experience. Only staining and light boric acid residue indicated the presence of a leak. Therefore the event had low safety significance.
Based on the above discussion, this condition does not constitute a Safety System Functional Failure as defined in Nuclear Energy Institute document 99-02 and there was no reduction in the public health and safety.
CAUSE
Based upon the review of the facts, material of construction, fabrication methods, characteristics of the leakage detected and Operating Experience, CR-3 concludes that primary water stress corrosion cracking (PWSCC) is the most likely failure mechanism for the pressurizer upper level instrument tap nozzle cracking. The use of Alloy 600/82/182 led to the development of PWSCC, which eventually grew to provide a path through which the reactor coolant was able to leak to the N RC FORM 366A (1.2001) outside of the pressurizer. This failure mode is consistent with the PWSCC failures in RCS nozzles of similar material, design, installation practice, and environment.
The Alloy 600/82/182 material is susceptible to PWSCC. High residual stresses created from the welding process, the environment created by the RCS water chemistry, and operating temperatures promote PWSCC. The amount of leakage observed is consistent with leakage seen at other Pressurized Water Reactors in the same general location (steam space).
CORRECTIVE ACTIONS
1. The three pressurizer upper level instrument tap nozzles were repaired using the "half- nozzle" technique (Engineering Change 51801 and Work Order 469645). The technique leaves the flaw in place, but moves the pressure boundary from the internal weld to an external location. Additionally, it replaces a segment of the Alloy 600 nozzle with Alloy 690, using a similar weld material (Alloy 52 or 52M), which are less susceptible to PWSCC at operating temperatures.
2. A review of Operating Experience suggested that leakage would be associated with the pressurizer steam space versus the water space. This was substantiated by no evidence of RCS leakage being found during bare metal inspections on any of the similar pressurizer nozzles: the lower level taps, RTD thermowell, water sample tap or vent nozzle.
3. Other actions (e.g., evaluating the need to revise the augmented Inservice Inspection program to bare metal inspect the pressurizer vent, lower level instrument taps, RTD thermowell, and water sample nozzle on some frequency) are being addressed in NCR 106443.
PREVIOUS SIMILAR EVENTS
This is the first occurrence of pressure boundary leakage of a pressurizer level instrument tap nozzle at CR-3. However, an occurrence of RCS pressure boundary leakage due to PWSCC of Alloy 600 material was reported to the NRC in Licensee Event Report 50-302101-004-00, "Reactor Pressure Vessel Head Leakage Due To Control Rod Drive Mechanism (CRDM) Nozzle Degradation," dated November 28, 2001.
ATTACHMENTS
Attachment 1 - Abbreviations, Definitions, and Acronyms Attachment 2 - List of Commitments The following table identifies those actions committed to by Progress Energy Florida, Inc., (PEF) in this document. Any other actions discussed in the submittal represent intended or planned actions by PEF. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Supervisor, Licensing & Regulatory Programs of any questions regarding this document or any associated regulatory commitments.
RESPONSE
SECTION
COMMITMENT DUE DATE
No regulatory commitments are being made in this submittal.