LER-2015-002, Regarding Indication in Control Rod Drive Mechanism Nozzle Weld Due to Embedded Flaws Opening Up from Thermal and Pressure Stresses During Operation |
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| Report date: |
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| 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 |
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| 4562015002R00 - NRC Website |
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NRCFORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 1/31/2017 (02-2014)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
13' PAGE Braidwood Station, Unit 1 05000456 1of3
~-TITLE Indication in Control Rod Drive Mechanism Nozzle Weld due to Embedded Flaws Opening Up from Thermal and Pressure Stresses during Operation
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NO.
MONTH DAY YEAR N/A N/A FACILITY NAME DOCKET NUMBER 04 03 2015 2015 - 002 -
00 06 02 2015 N/A N/A
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 6 D 20.2201(b)
D 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C)
D 50.73(a)(2)(vii)
D 20.2201 (ct)
D 20.2203(a)(3)(ii) 181 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) 0 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) 0 OTHER D 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B)
D 50. 73( a)(2)(v)(D)
Specify in Abstract below or in
O. Safety Consequences
This condition had no actual safety consequences impacting plant or public safety.
The flaw was identified in a timely manner and repaired. 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 embedded flaw repair to form a leak path through the reactor coolant pressure boundary.
Based on the A1R18 documented characteristics and dimensions of the observed PT indication, there was no Safety Significant functional failure (i.e., loss of safety function) as a result of these indications. The primary coolant pressure boundary was maintained and capable of preventing the release of radioactive material.
The Rod Drive system remained functional and there was no impact to structural integrity.
I:. Corrective Actions:
The identified indication was reduced to an acceptable dimension by mechanical means (manual grinding) and verified to be within the acceptance standard by liquid penetrant.
Develop and implement a corrective action plan that includes consideration of a long term Alloy 600 mitigation and other acceptable alternatives for the remaining acceptable, but unrepaired penetrations on the Braidwood Units' 1 and 2 RPV closure heads.
i:
Previous Occurrences
Previous Licensee Event Reports were made in April 2012 and November 2013 at Braidwood Station Unit 1 for indications on CROM penetration 69 (LER 2012-002-00 and LER 2013-002-00).
G. Component Failure Data
Manufacturer Westinghouse Nomenclature Reactor Vessel Integrated Head Package Termination 1718E72 Mfg. Part Number N/A
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| | | Reporting criterion |
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| 05000457/LER-2015-001, Regarding Steam Generator Atmospheric Relief Manual Isolation Valve Fails to Close During Quarterly Surveillance Due to Long Term Corrosion | Regarding Steam Generator Atmospheric Relief Manual Isolation Valve Fails to Close During Quarterly Surveillance Due to Long Term Corrosion | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
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)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(iii)
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(viii)(B) | | 05000456/LER-2015-001, Regarding Main Steam Isolation Valve Technical Specifications Applicability - MSIV Operability Position Change | Regarding Main Steam Isolation Valve Technical Specifications Applicability - MSIV Operability Position Change | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
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)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(viii)(B) | | 05000457/LER-2015-002, Regarding Automatic Actuation of Auxiliary Feedwater and Automatic Reactor Trip Signal Due to Startup Feedwater Pump Failing to Start on Demand and Motor Driven Feedwater Pump Elevated Bearing Temperature Exceeding | Regarding Automatic Actuation of Auxiliary Feedwater and Automatic Reactor Trip Signal Due to Startup Feedwater Pump Failing to Start on Demand and Motor Driven Feedwater Pump Elevated Bearing Temperature Exceeding | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
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)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(iii)
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(viii)(B) | | 05000456/LER-2015-002, Regarding Indication in Control Rod Drive Mechanism Nozzle Weld Due to Embedded Flaws Opening Up from Thermal and Pressure Stresses During Operation | Regarding Indication in Control Rod Drive Mechanism Nozzle Weld Due to Embedded Flaws Opening Up from Thermal and Pressure Stresses During Operation | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
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)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(iii)
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(viii)(B) | | 05000456/LER-2015-003, Regarding Unanalyzed Condition Due to a Design Deficiency with Pressurizer Power Operated Relief Valve Circuitry That Could Prevent Valve Manual Closure to Mitigate Spurious Operation | Regarding Unanalyzed Condition Due to a Design Deficiency with Pressurizer Power Operated Relief Valve Circuitry That Could Prevent Valve Manual Closure to Mitigate Spurious Operation | 10 CFR 50.73(a)(2)(iv)(A), System Actuation
10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
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)(v), Loss of Safety Function
10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
10 CFR 50.73(a)(2)(i)
10 CFR 50.73(a)(2)(x)
10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
10 CFR 50.73(a)(2)(iii)
10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
10 CFR 50.73(a)(2)(ix)(A), Prevented Safety Function in Multiple System
10 CFR 50.73(a)(2)(viii)(A)
10 CFR 50.73(a)(2)(viii)(B) | | 05000457/LER-2015-003, Regarding Degraded Valve Causes Loss of Diesel Fuel Oil System Volume for One Train of the Diesel Generator System Resulting in Inoperability Longer than Allowed by Technical Specifications | Regarding Degraded Valve Causes Loss of Diesel Fuel Oil System Volume for One Train of the Diesel Generator System Resulting in Inoperability Longer than Allowed by Technical Specifications | |
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