05000412/LER-2012-002
| Beaver Valley Power Station Unit Number 2 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded |
| Initial Reporting | |
| ENS 48387 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded |
| 4122012002R00 - NRC Website | |
Energy Industry Identification System (EllS) codes are identified in the text as [XX].
CONDITIONS PRIOR TO OCCURRENCE
Unit 2: Mode 6, Refueling.
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 October 6, 2012, during the Beaver Valley Power Station, Unit No. 2 (BVPS-2) refueling outage (2R16), the results of planned ultrasonic examinations performed on Penetration No. 44 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. 44 identified an indication as having a primary water stress corrosion cracking (PWSCC) flaw located at the top of the J groove weld at the 346.5 degree location. The flaw was in the axial direction with a length of 0.28 inches. A liquid penetrant examination was also performed to determine if this indication was open to the surface. The liquid penetrant examination identified a 0.5 inch long by 0.0625 inch wide linear indication at the 340 degree location of the penetration and confirmed that the indication was open to the surface.
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 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. 44 was repaired as required prior to plant startup from the 2R16 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. 44 was reported to the Nuclear Regulatory Commission (NRC) per 10 CFR 50.72 on October 6, 2012 (EN 48387).
Liquid penetrant examination also revealed two recordable indications on previous weld overlay repairs on Penetrations No.16 and 57 which were not within the RCS pressure boundary material. Each was a rounded indication on a weld overlay greater than 0.1875 inch. Rounded indications of greater than 0.1875 inch must be remediated per the NRC approved relief request for BVPS-2. The indication on Penetration No. 57 was removed by buffing. The indication on Penetration No. 16 was removed by grinding and weld material was applied to restore the seal weld thickness. These penetrations were then found acceptable in accordance with the applicable ASME code criteria.
CAUSE OF EVENT
Primary Water Stress Corrosion Cracking (PWSCC) of the Alloy 600 J groove weld 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. 44 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-2 reactor vessel upper head penetration indication found on Penetration No. 44 during the 2R16 refueling outage inspection is considered to be very low.
This is based on the estimated conditional core damage probability (CCDP) and conditional large early release probability (CLERP) for the event. For this analysis, since the indication was not through wall and there was no evidence of RCS leakage, a 5 percent probability that the weld flaws could have propagated and resulted in a reactor vessel head penetration Loss of Coolant Accident (LOCA) was conservatively assumed.
CORRECTIVE ACTIONS
1. Reactor head Penetration No. 44 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 2009, 2008, and 2006. The BVPS-2 reactor vessel head is an original component. BVPS Unit 1 replaced its reactor vessel head in 2006.
CR 2012-15729