05000400/LER-2015-003
| Shearon Harris Nuclear Power Plant, Unit 1 | |
| Event date: | 04-07-2015 |
|---|---|
| Report date: | 05-28-2015 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| 4002015003R00 - NRC Website | |
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
2. DOCKET
Energy Industry Identification System (EIIS) and component codes are identified in the text as [XX].
On April 7, 2015, and April 9, 2015, the Harris Nuclear Plant was shut down for a scheduled refueling outage in mode six, at 0% power. The reactor pressure vessel head [RPV] penetration nozzles [NZL] in the reactor coolant system [AB] were being examined as required by 10 CFR 50.55a(g)(6)(ii)(D). Ultrasonic examinations identified indications that required repair in head penetration nozzles 14, 18, and 23.
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 three indications that required repair. The indications were approximately 0.233, 0.260, and 0.297 inches long in nozzles 14, 18, and 23 respectively, and axial in orientation. The indications were repaired using the inside diameter temper bead welding process. The elapsed time from discovery on April 7 and 9 until the nozzles were repaired on April 30, 2015 was approximately 23 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 flaws in nozzle 14, 18, and 23 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
Nozzles 14, 18, and 23 were 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.
Safety Analysis
The ultrasonic testing results revealed that the flaws were 16%, 32%, and 23% through-wall extent in nozzles 14, 18, and 23 respectively, 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 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 safety- significant failures. 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
LERs 2013-001-00 and 2013-003-00 reported previously identified and repaired flaws on the Harris reactor vessel closure head nozzles. The reported indications in those LERs also exhibited characteristics of PWSCC.
This report contains no regulatory commitments.