05000412/LER-2014-001
| Beaver Valley Power Station Unit Number 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(C), 50.54(x) TS Deviation |
| Initial Reporting | |
| ENS 50079 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| 4122014001R00 - NRC Website | |
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
CONDITIONS PRIOR TO OCCURRENCE
Unit 2: Undefined Mode with no fuel in the reactor vessel.
There were no systems, structures, or components that were inoperable at the start of the event that contributed to the event.
DESCRIPTION OF EVENT
On May 1, 2014, during the Beaver Valley Power Station, Unit No. 2 (BVPS-2) seventeenth refueling outage (2R17), the results of planned ultrasonic examinations performed on Penetration No. 41 of the reactor vessel head [AB] did not meet the applicable acceptance criteria. The indication was not through wall as there was no evidence of leakage at this penetration based on inspections performed on top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) for reactor vessel head inspections to find potential flaws/indications well before the structural integrity of the reactor vessel head pressure boundary is significantly challenged.
The ultrasonic examination of Penetration No. 41 identified an indication as having a primary water stress corrosion cracking (PWSCC) flaw located on the downhill side of the penetration tube. The surface flaw was in the axial direction with a length of 0.56 inches.
The reactor vessel head examination is a requirement of 10 CFR 50.55a(g)(6)(ii)(D) which invokes ASME Code Case N-729-1. Ultrasonic examinations or eddy current examinations (as applicable) are performed on each of the 66 vessel head penetrations on the BVPS-2 head during each refueling outage until head replacement activities are completed in the future. Head Penetration No. 41 was repaired as required prior to plant startup from the 2R17 refueling outage.
Indications that cannot be dispositioned as acceptable per ASME Code Section XI in a Reactor Coolant System (RCS) pressure boundary are reportable under 10CFR 50.72(b)(3)(ii)(A) / 50.73(a)(2)(ii)(A) as a condition of the nuclear power plant, including its principal safety barriers, being degraded. The indication found on Penetration No. 41 was reported to the Nuclear Regulatory Commission (NRC) per 10 CFR 50.72 on May 1, 2014 (EN 50079).
Liquid penetrant examination also revealed two recordable indications on previous weld overlay repairs on Penetrations No. 44 and 57 which were not within the RCS pressure boundary material. Each was a rounded indication on a weld overlay of approximately 0.125 inch. Rounded indications of greater than 0.1875 inch must be remediated per the NRC approved relief request for BVPS-2. However, the indications on Penetrations No. 44 and No. 57 were removed by buffing. These penetrations were determined to be acceptable in accordance with the applicable ASME code criteria.
CAUSE OF EVENT
Primary Water Stress Corrosion Cracking (PWSCC) of the Alloy 600 penetration tube material was determined to be the apparent cause of the identified flaw. The failure mechanism is a known issue that is addressed by the requirements of 10 CFR 50.55a(g)(6)(ii)(D). The repair to Penetration No. 41 utilized an embedded flaw weld overlay to encapsulate the surface attached flaw with an Alloy 690 material. A weld overlay of PWSCC resistant material to isolate the indication from the borated water environment arrests any further PWSCC on the penetration. The embedded flaw weld overlay methodology was previously approved by the NRC for use at BVPS-2.
ANALYSIS OF EVENT
The safety significance associated with the BVPS Unit 2 reactor vessel upper head penetration indication found on Penetration 41 during the 2R17 refueling outage inspection is considered to be very low. This is based on the fact that the indication was not through wall, there was no indicated RCS leakage, and the estimated conditional core damage probability and conditional large early release probability for the event are very small, assuming that there is a 0.1 percent probability that the axial weld flaw could have propagated circumferentially to become a through-weld failure and ejected the penetration tube, resulting in a reactor vessel head penetration Loss of Coolant Accident (LOCA).
CORRECTIVE ACTIONS
1. Reactor head Penetration No. 41 was repaired in accordance with the applicable embedded flaw repair methodology approved by the NRC for use at BVPS-2.
Planning is ongoing for a reactor vessel head replacement which is expected to occur during a future outage at BVPS-2.
PREVIOUS SIMILAR EVENTS
Similar reactor vessel head indications were found and repaired during the previous BVPS-2 refueling outages in 2012, 2009, 2008, and 2006. The BVPS-2 reactor vessel head is an original component.
BVPS Unit 1 replaced its reactor vessel head in 2006 with a new head containing PWSCC resistant material.
CR 2014-08078