05000255/LER-2007-002

03-19-2007

minute) was identified in VHX-4 containment air cooler (CAC). The leak was subsequently determined to be from a pin-hole in a single cooling coil tube at the H-bend fitting.

The CAC cooling coil is a part of the containment boundary, and is required to be structurally sound to ensure that post accident containment leakage will not exceed allowable leakage (La). Therefore, Technical Specification Limiting Condition For Operation 3.6.1, "Containment," is initially applicable for any through-wall defect noted in a CAC cooling coil.

Although the size of the identified pin-hole would not have challenged La, it was not possible to characterize the flaw to verify the operability of the cooling coil for post-accident conditions over the CAC's mission time.

ihis event is reportable in accordance with 10 CFR 50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

LER-2007-002,
Docket Number
Event date:	01-19-2007
Report date:	03-19-2007
Reporting criterion:	10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material
2552007002R00 - NRC Website
v • d • e

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Palisades Nuclear Plant 05000-255 2007 -- 002 -- 00

EVENT DESCRIPTION

On January 19, 2007, with the plant in Model, a small service water [BI] leak (approximately 60 drops per minute) was identified in VHX-4 containment air cooler (CAC) [CLR;BK]. The leak was subsequently determined to be from a pin-hole in a single cooling coil tube at the H-bend fitting.

The CAC cooling coil is a part of the containment [NH] boundary, and is required to be structurally sound to ensure that post accident containment leakage will not exceed allowable leakage (La).

Therefore, Technical Specification (TS) Limiting Condition For Operation (LCO) 3.6.1, "Containment," is initially applicable for any through-wall defect noted in a CAC cooling coil.

Although the size of the identified pin-hole would not have challenged La, it was not possible to characterize the flaw using the criteria prescribed in American Society of Mechanical Engineers (ASME) Code Case N-513-1 to verify the operability of the cooling coil for post-accident conditions over the CAC's mission time.

Therefore, this event is reportable in accordance with 10 CFR 50.73(a)(2)(v)(C) as a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material.

Since VHX-4 has no safety-related cooling function, TS LCO 3.6.6, "Containment Cooling Systems" was not affected.

CAUSE OF THE EVENT

The probable cause of the through-wall flaw is loss of wall thickness due to flow induced erosion.

CORRECTIVE ACTIONS

VHX-4 was repaired using an ASME accepted method by plugging tubes to isolate the H-bend flaw.

Subsequently, VHX-4 was completely isolated by installing blanks on the service water inlet and individual cooling coil outlet flanges.

The VHX-4 CAC cooling coils are currently scheduled to be replaced during the Fall 2007 refueling outage.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Palisades Nuclear Plant 05000-255 2007 -- 002 -- 00

SAFETY SIGNIFICANCE

The safety significance of this occurrence is considered to be minimal. For a worst case failure of a single CAC cooling coil tube, analysis has determined that all on-site and off-site dose limits would continue to be met for the bounding postulated post-accident scenario.

PREVIOUS SIMILAR EVENTS

Licensee Event Report 05-006, "Inoperable Containment Due to Containment Air Cooler Through- Wall Flaws" Licensee Event Report 06-006, "Inoperable Containment Due to Containment Air Cooler Through- Wall Flaw" Licensee Event Report 06-008, "Inoperable Containment Due to Containment Air Cooler Through- Wall Flaw"

Licensee Event Report

Docket Number

70 | 71 | 72 | 73 | 74 | 75 | 76 | 77 | 78 | 79 | 80 | 81 | 82 | 83 | 84 | 85 | 86 | 87 | 88 | 89 |

90 | 91 | 92 | 93 | 94 | 95 | 96 | 97 | 98 | 99 | 00 | 01 | 02 | 03 | 04 | 05 | 06 | 07 | 08 | 09 |

10 | 11 | 12 | 13 | 14 | 15 | 16 | 17 | 18 | 19 | 20 |

		Reporting criterion
05000498/LER-2007-001	Turbine-Driven Auxiliary Feedwater Pump Failed to Start During Surveillance Testing (Supplement 1)	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000315/LER-2007-001	-f Unit 1 Automatic Reactor Trip	10 CFR 50.73(a)(2)(iv)(A), System Actuation
05000263/LER-2007-001
05000266/LER-2007-001		10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
05000269/LER-2007-001	Dual Unit Trip from Jocassee Breaker Failure	10 CFR 50.73(a)(2)(iv)(A), System Actuation
05000272/LER-2007-001	ESF Actuation of Auxiliary Feedwater Pumps in Mode 3.	10 CFR 50.73(a)(2)(iv)(A), System Actuation
05000265/LER-2007-001	Manual Reactor Scram on Increasing Condenser Backpressure Due to a Decrease in 2A Offgas Train Efficiency	10 CFR 50.73(a)(2)(iv)(A), System Actuation
05000278/LER-2007-001	Laboratory Analysis Identifies Safety Relief Valves and Safety Valve Set Point Deficiencies	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000249/LER-2007-001	Unit 3 High Pressure Coolant Injection System Declared Inoperable	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
05000282/LER-2007-001		10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000247/LER-2007-001	450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249Entergy Tel (914) 734-6700 Fred Dacimo Site Vice President Administration February 28, 2007 Indian Point Unit No. 2 Docket No. 50-247 NL-07-013 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001 Subject:L Licensee Event Report # 2007-001-00, "Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable Residual Heat Removal Pump Due to an Electrical Supply Breaker Failure" Dear Sir: Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2007-001-00. The enclosed LER identifies an event where the plant was operated in a condition prohibited by Technical Specifications, which is reportable under 10 CFR 50.73(a)(2)(i)(B). This condition has been recorded in the Entergy Corrective Action Program as Condition Report CR-IP2-2007-00013. There are no commitments contained in this letter. Should you or your staff have any questions regarding this matter, please contact Mr. Patric W. Conroy, Manager, Licensing, Indian Point Energy Center at (914) 734-6668. Sincerely, -Thr red R. Dacimo ite Vice President Indian Point Energy Center E Docket No. 50-247 NL-07-013 Page 2 of 2 Attachment: LER-2007-001-00 CC: Mr. Samuel J. Collins Regional Administrator — Region I U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident Inspector's Office Resident Inspector Indian Point Unit 2 Mr. Paul Eddy State of New York Public Service Commission INPO Record Center NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104DEXPIRES: 06/30/2007 (6-2004) Estimated burden per response to comply with this mandatory collection request: 50 hours.DReported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internetLICENSEE EVENT REPORT (LER) e-mail to infocollects@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-l0202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. 2. DOCKET NUMBER 1 3. PAGE1. FACILITY NAME: INDIAN POINT 2 05000-247 1 OF 4 4. TITLE: Technical Specification Prohibited Condition Due to Exceeding the Allowed Completion Time for an Inoperable Residual Heat Removal Pump Due to an Electrical Supply Breaker Failure	10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ix)
05000483/LER-2007-001	. Single Train Inoperability in the Essential Service Water System due to Inadequate Valve Closure Setup
05000286/LER-2007-001	450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249Entergy Tel (914) 734-6700 Fred Dacimo Site Vice President June 4, 2007 Indian Point 3 Docket No. 50-286 N L-07-052 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001 Subject:LLicensee Event Report # 2007-001-00, "Manual Reactor Trip Due to Decreasing Steam Generator Levels as a Result of the Loss of Feedwater Flow Caused by the Failure of 32 Main Feedwater Pump Train A Control Logic Power Supply" Dear Sir or Madam: Pursuant to 10 CFR 50.73(a)(1), Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2007-001-00. The attached LER identifies an event where the reactor was manually tripped while critical, which is reportable under 10 CFR 50.73(a)(2)(iv)(A) . This condition has been recorded in the Entergy Corrective Action Program as Condition Report CR-IP3-2007-01775. There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. T. R. Jones, Manager, Licensing at (914) 734-6670. Sincerely, Fred R. Dacimo Site Vice President Indian Point Energy Center cc:LMr. Samuel J Collins, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 3 Mr. Paul Eddy, New York State Public Service Commission INPO Record Center pP,c.1)-1 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES: 6/30/2007 (6-2004) Estimated burden per response to comply with this mandatory collection request:D50 hours.DReportedDlessons learned areDincorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internetLICENSEE EVENT REPORT (LER) e-mail to infocollects@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. 1. FACILITY NAME INDIAN POINT 3 2. DOCKET NUMBER 13. PAGE 05000-286 1 OFTD5 4. TITLE Manual Reactor Trip Due to Decreasing Steam Generator Levels as a Result of the Loss of Feedwater Flow Caused by the Failure of 32 Main Feedwater Pump Train A Control Logic Power Supply	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function
05000293/LER-2007-001		10 CFR 50.73(a)(2)(iv)(A), System Actuation
05000306/LER-2007-001		10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function
05000309/LER-2007-001	Uncompensated Degradation in a Security System
05000414/LER-2007-001	Failure to Comply with Action Statement in Technical Specification (TS) 3.3.1 for Loss of a Channel of the Solid State Protection System	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000311/LER-2007-001	Inoperability of the Chilled Water System - (21 and 22 Chillers Inoperable)	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000331/LER-2007-011	. Undervoltage ConditiOn Resulted in the Actuation of the Emergency Diesel Generators
05000346/LER-2007-001	Station Vent Radiation Monitor in Bypass due to Faulty Optical Isolation Board	10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000348/LER-2007-001	Vire President - Farley Operating Company, Inc. Po51 Office Drawer 470 Ashford, Alabarid 36312-0470 Tel 334 814 4511 Fax 334 814 4728 SOUTHERN June 22, 2007 COMPANY Energy to Serve Your World Docket Nos.: 50-348 NL-07-1231 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant — Units 1 and 2 Licensee Event Report 2007-001-00 Technical Specification 3.8.1 Violation Due to Failure of Breaker / Mechanism-Operated Cell Switch Ladies and Gentlemen: Joseph M. Farley Nuclear Plant - Licensee Event Report (LER) No. 2007-001-00 is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B). This letter contains no NRC commitments. If you have any questions, please advise. Sincerely, 7e. R. Johnson Vice President — Farley Joseph M. Farley Nuclear Plant 7388 North State Highway 95 Columbia AL 36319 JRJ/CHM Enclosure: Licensee Event Report 2007-001-00 - Unit 1 U. S. Nuclear regulatory Commission NL-07-1231 Page 2 cc:� Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President — Farley Mr. D. H. Jones, Vice President — Engineering RTYPE: CFA04.054; LC # 14596 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Ms. K. R. Cotton, NRR Project Manager — Farley Mr. E. L. Crowe, Senior Resident Inspector— Farley NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007 (6-2004) Estimated burden per response to comply with this mandatory collection request: 50 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nudear Regulatory Commission, Washington, DC 2055570001, or by InternetLICENSEE EVENT REPORT (LER) e-mail to infocolledsanrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose an information(See reverse for required number of collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, thedigits/characters for each block) information collection. 1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE Joseph M. Farley Nuclear Plant - Unit 1 05000 348 1 OF 4 4. TITLE Technical Specification 3.8.1 Violation Due to Failure of Breaker / Mechanism-Operated Cell (MOC) Switch	10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000316/LER-2007-001	As-Found Local Leak Rate Tests Not Performed	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000456/LER-2007-001	Unit 1 Reactor Trip Following a 345 Kv Transmission Line Lightning Strike	10 CFR 50.73(a)(2)(iv)(A), System Actuation
05000333/LER-2007-001		10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000254/LER-2007-001	Quad Cities Nuclear Power Station Unit 1 05000254 1 of 3	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000389/LER-2007-001	S, Reactor Shutdown Due to Unidentified RCS Leakage	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
05000255/LER-2007-001		10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000369/LER-2007-001	369 5McGuire Nuclear Station Unit 1 05000 1 OF5
05000335/LER-2007-001	Mispositioned Service Air Containment Isolation Valves	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
05000362/LER-2007-001	Failure to declare Emergency Diesel Generator Inoperable and enter TS Action	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000353/LER-2007-001	Scram Discharge Volume Vent and Drain Valves Opened Due To Fuse Removal	10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability
05000400/LER-2007-001	Control Rod Shutdown Bank Anomaly Causes Entry into Technical Specification 3.0.3	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000261/LER-2007-001	Reactor Trip Due to a Loose Wire in the Main Transformer Monitoring Circuitry	10 CFR 50.73(a)(2)(iv)(A), System Actuation
05000389/LER-2007-002	2B2 Reactor Coolant Pump (RCP) Seal Housing Leakage	10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
05000255/LER-2007-002		10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material
05000395/LER-2007-002	Failure to Follow Administrative Controls Results in LCO 3.6.4 Violation	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000306/LER-2007-002		10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000440/LER-2007-002	Shutdown Cooling Pump Trip Results in Operation Prohibited by Technical Specifications	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i)
05000414/LER-2007-002	Technical Specification Violation Associated with Containment Valve Injection Water System	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000416/LER-2007-002	Reactor SCRAM due to Turbine Trip caused by Loss of Condenser Vacuum	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv), System Actuation
05000423/LER-2007-002	Loss of Offsite Power Caused by Transmission System Operator While Defueled	10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv), System Actuation
05000311/LER-2007-002	RReactor Trip Due to a Breach in the Condensate System	10 CFR 50.73(a)(2)(iv)(A), System Actuation
05000369/LER-2007-002
05000454/LER-2007-002	Technical Specification Required Shutdown of Unit 1 and Unit 2 Due to an Ultimate Heat Sink Pipe Leak Common to Both Units
05000282/LER-2007-002		10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(ii)
05000315/LER-2007-002	Failure to Declare Essential Service Water Inoperable	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000247/LER-2007-002	Technical Specification Prohibited Condition Due to Exceeding Containment Air Temperature Limit Allowed Outage Time as a Result of Changes in Instrument Uncertainty	10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
05000250/LER-2007-002	Completion of Shutdown Required by Technical Specifications due to Inoperable Rod Position Indication for Two Control Rods in the Same Control Bank	10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown
05000353/LER-2007-002	Automatic Actuation of Main Condenser Low Vacuum Isolation Logic During Refueling Outage	10 CFR 50.73(a)(2)(iv), System Actuation
05000272/LER-2007-002	MManual Reactor Trips Due to Degraded Condenser Heat Removal	10 CFR 50.73(a)(2)(iv)(A), System Actuation
... further results

05000255/LER-2007-002

Contents

EVENT DESCRIPTION

CAUSE OF THE EVENT

CORRECTIVE ACTIONS

SAFETY SIGNIFICANCE

PREVIOUS SIMILAR EVENTS

Navigation menu

05000255/LER-2007-002

EVENT DESCRIPTION

CAUSE OF THE EVENT

CORRECTIVE ACTIONS

SAFETY SIGNIFICANCE

PREVIOUS SIMILAR EVENTS

Navigation menu

Search