05000456/LER-2012-002, Regarding Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Attributed to Primary Water Stress Corrosion Cracking
| ML12174A227 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 06/22/2012 |
| From: | Enright D Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BW120061 LER 12-002-00 | |
| Download: ML12174A227 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 4562012002R00 - NRC Website | |
text
Exelon Generation Company, LLC Braidwood Station 35100 South Route 53, Suite 84 Braceville,IL 60407*9619 June 22, 2012 BW120061 www.exeloncorp.com 10 CFR 50.73 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Facility Operating License No. NPF-72 NRC Docket No. STN 50-456
Subject:
Licensee Event Report 2012-002 Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Attributed to Primary Water Stress Corrosion Cracking The enclosed Licensee Event Report (LER) is being submitted in accordance with 10 CFR 50.73, "Licensee Event Report System."
There are no regulatory commitments contained in this letter. Should you have any questions concerning this submittal, please contact Mr. Chris VanDenburgh, Regulatory Assurance Manager, at (815) 417-2800.
Respectfully, Daniel J. Enright Site Vice President Braidwood Station
Enclosure:
LER 2012-002-00 cc: NRR Project Manager - Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety US NRC Regional Administrator, Region III US NRC Senior Resident Inspector (Braidwood Station)
Illinois Emergency Management Agency - Braidwood Representative
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
- 13. PAGE Braidwood Station, Unit 1 05000456 1 of 3
- 4. TITLE Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Attributed to Primary Water Stress Corrosion Cracking
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR I SEQUENTIAL I REV MONTH DAY YEAR N/A N/A NUMBER NO.
FACILITY NAME DOCKET NUMBER 04 23 2012 2012 - 002 - 00 06 22 2012 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 20.2201 (b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(i)(C)
D 50.73(a)(2)(vii)
N/A Defueled D 20.2201 (d)
D 20.2203(a)(3)(ii)
~ 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A)
D 20.2203(a)(1)
D 20.2203(a)(4)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A)
D 50.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
D 20.2203(a)(2)(iii)
D 50.36(c)(2)
D 50.73(a)(2)(v)(A)
D 73.71(a)(4) 000 D 20.2203(a)(2)(iv)
D 50.46(a)(3)(ii)
D 50.73(a)(2)(v)(B)
D 73.71 (a)(5)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(C)
D OTHER D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(v)(D)
Specify in Abstract below or in This was the second Volumetric Examination performed on the Braidwood Unit 1 reactor head penetrations. The first Volumetric Examination was performed during the April 2006 refueling outage. No evidence of PWSCC was identified during the 2006 examination.
Prior to the Unit 1 refueling outage, the Volumetric Examination was every four outages and the frequency for the Bare Metal Visual Examination was every three outages. As a result of discovery of the indication in Penetration 69, both examinations are now required every refuel outage.
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Safety Consequences
This condition had no actual safety consequences impacting plant or pUblic safety. The flaw was identified in a timely manner and repaired prior to through wall leakage occurring. The flaw was identified as part of a required periodic inspection. Potentially, if the flaw remained undetected, it could have over time propagated through the Alloy 600 weld material to form a leak path through the reactor coolant pressure boundary.
Based on the Unit 1 spring 2012 documented characteristics and dimensions of the flaw (axially oriented with a linear extent of 0.600 inches and a through wall depth approximately 33.5% through wall), there was no safety significant functional failure as a result of this flaw as no safety functions were lost. The primary coolant pressure boundary was maintained and capable of preventing the release of radioactive material. The rod drive system remained functional.
10..
Corrective Actions
The indication in Penetration 69 was repaired prior to the return to service of the Unit 1 reactor head.
The frequency of examination of Unit 1 for both the Bare Metal Visual Exam and the Volumetric Exam has been changed to every refueling outage.
Previous Occurrences
No previous, similar Licensee Event Reports were identified at the Braidwood Station.
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Component Failure Data
Manufacturer N/ANomenclature N/A Model N/A Mfg. Part Number N/A