ENS 47934
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ENS Event | |
|---|---|
06:00 Apr 23, 2012 | |
| Title | Ultrasonic Examination Results in an Indication on the Reactor Pressure Vessel Head Penetration |
| Event Description | On April 23, 2012, during the Braidwood Station Unit 1 refueling outage, it was determined that the results of planned ultrasonic (UT) examinations performed on one penetration (Penetration No. 69) did not meet the applicable acceptance criteria, requiring repair prior to returning the reactor pressure vessel head to service. A portion of the indication was conservatively assumed to be within the J-groove weld. At the time of the examination, the Braidwood Station Unit 1 reactor pressure vessel head was classified as a low susceptibility head. The cause of the recordable indication is attributed to Primary Water Stress Corrosion Cracking. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to ensure the structural integrity of the reactor pressure vessel head pressure boundary. All of the remaining penetrations have been examined during the current refueling outage. Repairs to Penetration No. 69 were completed prior to commencing startup. No other repairs were required.
This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as found indication did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. The NRC Resident Inspector has been notified. As a result of a follow-up review, at 0600 CDT on 05/18/12, the licensee determined this was an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reportable event while Unit 1 was in Mode 4. The licensee did not determine this was a reportable event immediately after the UT was conducted on April 23, 2012. The unit is currently in Mode 3. Reactor Vessel Head Penetration No. 69 is a control rod penetration. Reactor Coolant Leakage test are satisfactory and there are no indications of boron buildup on the Reactor Vessel Head. Control rod testing has not yet been performed. |
| Where | |
|---|---|
Braidwood  Illinois (NRC Region 3) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | |
| Time - Person (Reporting Time:+606.83 h25.285 days <br />3.612 weeks <br />0.831 months <br />) | |
| Opened: | David Kortge 12:50 May 18, 2012 |
| NRC Officer: | John Shoemaker |
| Last Updated: | May 18, 2012 |
| 47934 - NRC Website
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WEEKMONTHYEARENS 579782025-10-10T07:30:00010 October 2025 07:30:00
[Table view]10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Degraded Control Rod Drive Mechanism Penetration ENS 561532022-10-07T06:19:0007 October 2022 06:19:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Control Rod Drive Mechanism (CRDM) Penetration Degraded ENS 536522018-10-08T05:00:0008 October 2018 05:00:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Reactor Coolant System Pressure Boundary Leakage ENS 533542018-04-20T22:30:00020 April 2018 22:30:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Degraded Reactor Vessel Head ENS 522752016-10-02T21:00:0002 October 2016 21:00:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Liquid Penetration Examination Results in Indications on Reactor Vessel Head Penetration ENS 509532015-04-03T21:31:0003 April 2015 21:31:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Liquid Penetration Examination Results in Indication on Reactor Vessel Head Penetration ENS 493532013-09-18T10:10:00018 September 2013 10:10:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Boroscopic Inspection of Reactor Vessel Head Determined Evidence of Boundary Leakage ENS 493432013-09-14T07:00:00014 September 2013 07:00:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Liquid Penetration Examination Results in Indication on Reactor Vessel Head Penetration ENS 479342012-04-23T06:00:00023 April 2012 06:00:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Ultrasonic Examination Results in an Indication on the Reactor Pressure Vessel Head Penetration 2025-10-10T07:30:00 | |
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