05000400/LER-2013-003
| Shearon Harris Nuclear Power Plant, Unit 1 | |
| Event date: | 11-18-2013 |
|---|---|
| Report date: | 01-15-2014 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(a) |
| 4002013003R00 - NRC Website | |
Background
Energy Industry Identification System (EllS) codes are identified in the text as [XC].
On November 18, 2013, the Harris Nuclear Plant was shut down for a scheduled refueling outage in mode six, at 0% power. The reactor pressure vessel head penetration nozzles [RPV-NZL] in the reactor coolant system [AB] were being examined as required by 10 CFR 50.55a(g)(6)(ii)(D). Ultrasonic examinations identified an indication in head penetration nozzle 37.
There were no systems, structures, or components that were inoperable at the start of the event that contributed to the event.
This condition is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(a), as an event or condition that resulted in the condition of the nuclear power plant, including its principal barriers, being degraded.
Event Description
Ultrasonic test data revealed an indication approximately 0.46 inches long and axial in orientation. The indication was repaired using the inside diameter temper bead welding process. The elapsed time from discovery on November 18 until the nozzle was repaired on December 2 was approximately 14 days.
The remaining control rod drive mechanism nozzles were also examined using nondestructive methods, and a surface examination of the vent line was performed. A bare metal visual examination of the top of the reactor vessel closure head was completed with no indications of leakage.
The reactor pressure vessel closure head was manufactured by Chicago Bridge and Iron, Serial Number T40.
Causal Factors The cause of the flaw in nozzle 37 was attributed to Primary Water Stress Corrosion Cracking (PWSCC). PWSCC occurs under conditions of high tensile stresses (either operating or residual), conducive environment (temperature and chemistry), and susceptible material. There is widespread industry operating experience that documents PWSCC of Alloy 600 dissimilar metal weld configurations.
Corrective Actions
Nozzle 37 was repaired utilizing the inside diameter temper bead welding process. Per the requirement of 10 CFR 50.55a(g)(6)(ii)(D)(5), if flaws attributed to PWSCC have been identified, examinations are required to be performed on the reactor vessel head every refueling outage to identify flaws and ensure appropriate repairs are performed.
Shearon Harris Nuclear Power Plant, Unit 1 05000400
Safety Analysis
The ultrasonic testing results revealed that the flaw was 21% through-wall and axial in orientation. An inspection of the exterior surfaces of the reactor head confirmed there was no leakage. The safety significance of the flaw's presence during operation was minimal. An extensive industry safety assessment of PWSCC in reactor vessel head penetrations concluded that a program of periodic nonvisual non-destructive examinations at appropriate intervals supplemented by periodic bare metal visual examinations provides adequate protection against potential safety-significant failures. Based on the industry safety assessment, it is reasonable to conclude that an inspection program in accordance with the requirements of ASME Code Case 729-1 as modified by the additional limitations set forth in 10 CFR 50.55a(g)(6)(ii)(D), provide assurance against any credible PWSCC degradation event that would challenge nuclear safety.
Additional Information
closure head nozzles. The reported indications in April 2012 and May 2013 also exhibited characteristics of PWSCC.
As stated previously, Energy Industry Identification System (EllS) codes are identified in the text as p0q.
This report contains no regulatory commitments.