05000339/LER-2001-003

LER-2001-003,

Document Number
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
3392001003R01 - NRC Website

LER NUMBER (6) FACILITY NAME (1)

DOCKET

05000 - 339 1.0 DESCRIPTION OF THE EVENT On October 28, 2001, North Anna Unit 2 was shutdown to perform a qualified, visual barehead inspection of the reactor pressure vessel (EIIS-RPV) head penetrations (EDS- PEN) for evidence of leakage as required by NRC Bulletin 2001-01, Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles. On October 31, 2001, with Unit 2 in Mode 5, this inspection resulted in three penetrations (51, 62 and 63) being identified as rejectable due to existence of boric acid on the reactor pressure vessel barehead surface. The other 62 penetrations were determined to be acceptable based on this inspection. On November 13, 2001, with Unit 2 in Mode 6, further review identified an apparent through-wall leak on penetration number 63 due to evidence of boric acid deposited at the base of the penetration and a liquid penetrant indication near the J- groove weld for that penetration. A portion of the J-groove weld, approximately two and three-fourths (2 3/4) inches long, containing the indication, was excavated to a depth of approximately one (1) inch. A liquid penetrant examination of this excavation on penetration 63 showed that the indication located in the outside edge of the weld almost the full length of the excavation remained and that it turned into the weld at the uphill and downhill ends of the excavation. Based on the apparent through-wall leak, a non- emergency 8-hour notification was made to the NRC, at 0859 hours0.00994 days <br />0.239 hours <br />0.00142 weeks <br />3.268495e-4 months <br />, on November 13, 2001, in accordance with 10CFR50.72(b)(3)(ii)(A), any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. This event is reportable per 10 CFR 50.73(a)(2)(ii)(A), for any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

In addition, Technical Specification (TS) 3.4.6.2 prohibits reactor coolant system (RCS) (EllS-AB) pressure boundary leakage in Modes 1 through 4. Although the apparent leakage was identified in Mode 6, it is reasonable to assume that the leakage occurred during Modes 1 through 4. Therefore, this event is also reportable in accordance with 10CFR50.73(a)(2)(i)(B) for a condition prohibited by TS.

A summary of NDE inspection results for penetration 63 and penetrations 51 and 62 that also had rejectable indications due to existence of boric acid on the reactor barehead surface is provided below.

Penetration # N2-63:

LP: � Liquid Penetrant (LP) examination of the entire weld surface revealed twelve indications at the toe of the weld, which appeared to be contained in the cladding material. No indications were noted on the weld surface. Six of the recorded indications were parallel to the weld, and six were transverse.

E/C ID: The thermal sleeve was removed to facilitate inspection of the tube. Eddy Current (E/C) inspection on the inside diameter (ID) of the penetration in the area FACILITY NAME (1)

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05000 - 339 (6)

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of the attachment weld under the vessel head was performed, and one 31 millimeter (mm) long axial indication was reported.

UT ID: A Ultrasonic Test (UT) examination of the ID was performed to confirm and determine the depth of the indication found using E/C. The depth of the indication was UT OD: A UT examination using pulse echo probes inside the tube (45° forward/up and 45° backward/down) was performed focusing on the outside diameter (OD) of the tube. There were no reportable indications using this method.

Penetration #N2-51:

LP: LP examination of the entire weld surface revealed twelve indications at the toe of the weld, which appeared to be contained in the cladding material. No indications were noted on the weld surface. Five of the indications were parallel to the weld, and seven were transverse.

E/C ID: This thermocouple penetration does not have a thermal sleeve. E/C inspection on the ID of the penetration in the area of the attachment weld under the vessel head was performed, and six axial indications were reported.

UT ID: UT examination of the ID of the tube was performed to confirm and determine the depth of the indications found using E/C. All indications were the exception of #3, that had a depth of determined using UT:

Indications #1 and #2, 24 mm long cluster (craze cracking) Indication #3, 10 mm long Indication #4, 6 mm long Indication #5, 12 mm long Indication #6, 6 mm long UT OD: A UT examination using pulse echo probes inside the tube (45° forward/up and 45° backward/down) was performed focusing on the OD of the tube. There were no reportable indications using this method.

Penetration # N2-62:

LP: LP examination of the entire weld surface revealed eight indications at the toe of the weld, which appeared to be contained in the cladding material. No indications were noted on the weld surface. Two of the recorded indications were parallel to the weld, and six were transverse.

E/C ID: The thermal sleeve was removed to facilitate inspection of the tube. E/C inspection on the inside diameter of the penetration in the area of the attachment weld under the vessel head was performed, and two axial cluster indications (craze cracking) were reported.

UT ID: A UT examination of the ID was performed to confirm and determine the depth of the indications found using E/C. Indication #1 was was Indication #1, 74 mm long Indication #2, 42 mm long UT OD: A UT examination using pulse echo probes inside the tube (45° forward/up and 45° backward/down) was performed focusing on the OD of the tube. There were no reportable indications using this method.

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS Reactor Coolant System leakage, including unidentified leakage, is quantified every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per TS 3.4.6.2. Prior to the Unit 2 shutdown to perform the reactor vessel head inspection, unidentified leakage was measured at 0.04 gpm and containment sump inleakage was measured at 0.06 gpm.

WCAP-14552, "Structural Evaluation of Reactor Vessel Upper Head Penetrations to Support Continued Operation: North Anna and Surry Units," was prepared to support determination of appropriate corrective actions. The WCAP documents that as much as 83.9% of the weld may be unfused, and the allowable stress limits can still be met. Even a complete lack of fusion in the zone between the weld and the head would not result in rod ejection because the weld to the tube would prevent it. Therefore, catastrophic failure of a penetration is unlikely. The health and safety of the public were not affected at any time during this event.

3.0 CAUSE The root cause of the reportable PT indication was hot-short cracking, which occurred during original fabrication of the reactor vessel head. The hot-short cracking was due to the presence of low melting point constituents in the weld. The reactor vessel head may not have been sufficiently cleaned prior to welding in the area around penetrations # 51, 62 and 63, which would indicate a lapse in process control. The hot-short cracking was not detected during the final PT of the finished J groove weld, because the indications were likely subsurface prior to startup. The cracking probably did not extend to the wetted weld surface until sometime later in North Anna Unit 2 reactor vessel head's operating

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05000 - 339 life. Alloy 82/182 welds are prone to cracking when subject to stress at elevated temperatures.

4.0 IMMEDIATE CORRECTIVE ACTION(S) Based on the qualified, visual barehead inspection results, additional NDE examinations were initiated to characterize the nature of the indications.

5.0 ADDITIONAL CORRECTIVE ACTIONS Evaluations of the eddy current and ultrasonic inspection results for the vessel head penetrations were performed to demonstrate that there is no evidence of a flaw propagating from the OD of the penetration or the penetration to weld fusion zone toward the ID and around the penetration circumferentially. The flaws in the affected penetrations were initially evaluated and reported using WCAP-14552 methodology, which included an assumed flaw aspect ratio of 6:1. However, some of the flaws found at North Anna had aspect ratios larger than 6:1. Subsequently, additional curves were prepared for flaw ratios of 15:1, 20:1, 30:1, 65:1, and 100:1. Using the appropriate bounding aspect ratio, the flaws were reanalyzed for North Anna Units 1 and 2. The revised remaining service life for Unit 2 penetration 51 is 4.16 Y, for Unit 2 penetration 62 is 3.82 Y and for Unit 2 penetration 63 is 4.70 Y.

A boat sample of the weld material from penetration 62 at a location where indications were identified by the LP exam was removed and sent to Westinghouse for laboratory analysis. This sample is considered representative of the indications found in the welds for penetrations 51 and 63 because they all are similarly located at the junction of the J- groove weld and the butter layer. Also, all of the indications are circumferential in nature with a similar appearance from LP. Analysis of the sample confirmed that the indications found using LP were most likely associated with original fabrication.

Relief requests were submitted to the NRC to allow alternative repair techniques for reactor vessel head penetrations that exhibited flaws. During a phone call with the NRC on November 16, 2001, North Anna was given verbal approval to use these repair techniques.

Repairs to the three vessel head penetration welds were performed in accordance with the above mentioned, NRC verbally approved relief requests.

Additional information related to the structural integrity of the reactor pressure vessel head penetration nozzles, including the extent of the leakage and indications, and the inspections and repairs undertaken to satisfy regulatory requirements will be provided in the written response to NRC Bulletin 2001-01.

LER NUMBER (6) DOCKET FACILITY NAME (1) 05000 = 339 North Anna Unit 2 was returned to service on Saturday, December 15, 2001.

6.0 ACTIONS TO PREVENT RECURRENCE Unit 2 was shutdown on September 8, 2002 for a scheduled refueling outage. Following reactor vessel head inspections the decision was made to replace the existing reactor vessel head. A new crack-resistant reactor vessel head was successfully installed and the unit was retumed to service on February 2, 2003.

7.0 SIMILAR EVENTS entering the "B" RCP thermal barrier housing. This line is part of the RCS pressure boundary leakage.

line for the "B" cold leg stop valve, which exceeded the TS Limit for RCS pressure boundary leakage.

8.0 MANUFACTURER/MODEL NUMBER Rotterdam Dockyard Company/Serial Number 30662 9.0 ADDITIONAL INFORMATION North Anna Unit 1 performed reactor pressure vessel head penetration inspections during the Fall 2001 refueling outage and no through-wall leaks were identified.

NORTH ANNA POWER STATION 2001 -003 01