05000412/LER-2009-002

LER-2009-002, Unacceptable Indications Identified During Reactor Vessel Head Inspection

Docket Number
Event date:	10-23-2009
Report date:	12-21-2009
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
Initial Reporting
ENS 45463	10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded
4122009002R00 - NRC Website

Energy Industry Identification System (ENS) codes are identified in the text as [XX].

DESCRIPTION OF EVENT

On October 23, 2009, during the Beaver Valley Power Station (BVPS) Unit No. 2 refueling outage (2R14), it was determined that the results of planned ultrasonic (UT) examinations performed on two penetrations of the reactor vessel head [AB] would not meet the applicable acceptance criteria. The indications are not through wall and there was no evidence of leakage 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 they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary.

Specifically, one circumferential indication was identified on penetration No. 57 at or near the toe of the J-groove weld between 352 and 5 degrees (approximately 0.45 inch arc length) and a maximum depth of 0.27 inch. A second circumferential indication was identified at penetration No. 49 at or near the toe of the J-groove weld between 39 and 60 degrees (approximately 0.75 inch arc length) and a maximum depth of 0.402 inch.

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 Unit 2 head during each refueling outage until head replacement activities are completed in the future. Both head penetration No. 49 and 57 were repaired as required prior to plant startup from 2R14 refueling outage.

Pursuant to Section 3.2.4 in NUREG 1022, Rev. 2, "Event Reporting Guidelines 10 CFR 50.72 and 50.73," indications that cannot be dispositioned as acceptable per ASME Code Section XI in a Reactor Coolant System 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. This was reported to the Nuclear Regulatory Commission per 10 CFR 50.72 on October 23, 2009 (EN Number 45463).

CAUSE OF EVENT

Primary Water Stress Corrosion Cracking (PWSCC) of the Alloy 600 J groove weld is the basic cause of the identified failure. The failure mechanism is a known issue to the industry, which is addressed by the requirements of 10 CFR 50.55a(g)(6)(ii)(D). The repairs to penetrations No. 49 and 57 utilize an embedded flaw weld overlay to encapsulate the surface attached flaw with an Alloy 690 material. By isolation of the Reactor Coolant CAUSE OF EVENT (Continued) System [AB] fluid from the crack, the further propagation of the flaw is eliminated. The embedded flaw weld overlay methodology was approved by the NRC (relief requests 2-TYP-3-RV-01 and 2-TYP-3-RV-03).

ANALYSIS OF EVENT

The safety significance associated with the BVPS Unit NO. 2 reactor vessel upper head penetration indications found on Penetrations No. 49 and 57 during the 2R14 refueling outage inspection is considered to be very low. This is,based on the estimated conditional core damage probability and Conditional large early releaSe probability for the event. Since the indications are not through wall and there was no evidence of RCS leakage, it is conservatively assumed there is a 5 percent probability that the weld flaws could have propagated and resulted in a reactor vessel head penetration Loss of Coolant Accident (LOCA).

CORRECTIVE ACTIONS

1. Reactor head penetrations No. 49 and 57 were repaired in accordance with the applicable embedded flaw repair methodology approved by the NRC.

Planning is underway for a reactor vessel head replacement whiCh is expected to occur during a future steam generator replacement outage.

PREVIOUS SIMILAR EVENTS

Similar reactor vessel head indications were found and repaired during the previous BVPS Unit 2 refueling outages in 2008 and 2006. The BVPS Unit 2 reactor vessel head is an original component. BVPS Unit 1 replaced its reactor vessel head in 2006 during its steam generator replacement outage.

CR 09-66489/66557