02-19-2008 | On December 21, 2007, at approximately 0825 hours0.00955 days <br />0.229 hours <br />0.00136 weeks <br />3.139125e-4 months <br />, with the plant operating in Mode 1 (Power Operation), at approximately 100 percent power, low flow occurred in the Annulus Exhaust Gas Treatment System ( AEGTS) train "B" while train "A" was out of service., The low flow resulted in less than the minimum allowed secondary containment annulus differential pressure. This condition was determined to be reportable in accordance with 10CFR50.73(a)(2)(v)(C) and (D), Loss of Safety Function. Additionally, this condition resulted in entry into Technical Specification Limiting Condition for Operation 3.0.3. Failure to correct this condition within one hour is reportable in accordance with 10CFR50.73(a)(2)(i)(B), a Condition Prohibited by Technical, Specifications.
The cause of the low flow was the failure of the discharge damper on the "B" Train. The.discharge damper failure was due to side-load wear induced binding of the shaft and shaft extension of the associated Hydramotor actuator. Both the discharge and the recirculation damper actuators were replaced. Maintenance instructions will be revised to ensure proper alignment of the shaft and shaft extension and proper return spring seating. Causal analysis will be performed on any future critical Hydramotor failures.
Since AEGTS is not a core damage mitigation system and does not mitigate Large and Early Containment releases, the unavailability of the AEGTS was determined to be of very low safety significance.
NRC FORM 366 (9-2007) PRINTED ON RECYCLED PAPER |
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LER-2007-006, Loss of Safety Function and Condition Prohibited by Technical Specifications due to Annulus Exhaust Gas Treatment System InoperabilityDocket Number |
Event date: |
12-21-2007 |
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Report date: |
02-19-2008 |
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Reporting criterion: |
10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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4402007006R00 - NRC Website |
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Energy Industry Identification System Codes are identified in the text as [XX].
I. INTRODUCTION
On December 21, 2007, at approximately 0825 hours0.00955 days <br />0.229 hours <br />0.00136 weeks <br />3.139125e-4 months <br />, with the plant operating in Mode 1 (Power Operation), at approximately 100 percent power, low flow occurred in the Annulus Exhaust Gas Treatment System (AEGIS) [BH] train "B" while train "A" was out of service. The low flow resulted in less than the minimum allowed secondary containment annulus differential pressure. This condition was determined to be reportable in accordance with 10CFR50.72(b)(3)(v)(C) and (D) as a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) control the release of radioactive material; or (D) Mitigate the consequences of an accident. Event Notification Form 43860 documents this report.
This event is being reported in accordance with 10CFR50.73(a)(2)(v)(C) and (D), an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (C) Control the release of radioactive material; or (D) Mitigate the consequences of and accident. It was later determined that both train "A" and "B" were inoperable from approximately 0826 hours0.00956 days <br />0.229 hours <br />0.00137 weeks <br />3.14293e-4 months <br /> until approximately 1012 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.85066e-4 months <br />, when train "A" was made operable. The unit was in Technical Specification (T.S.) Limiting Condition for Operation (LCO) 3.0.3 for a period of one hour and 46 minutes. Failure to correct this condition within one hour is reportable in accordance with 10CFR50.73(a)(2)(i)(B), a condition prohibited by Technical Specifications
II. EVENT DESCRIPTION
On December 21, 2007, AEGTS train "A " was removed from service to obtain a charcoal sample. At approximately 0825 hours0.00955 days <br />0.229 hours <br />0.00136 weeks <br />3.139125e-4 months <br />, the charcoal plenum for train "A" was opened to access the charcoal bed and obtain a sample. Opening the plenum resulted in a train "B" low flow alarm and a containment annulus (secondary containment) low differential pressure alarm when annulus differential pressure decreased to about 0.3 inches of vacuum water gauge. The minimum allowed annulus differential pressure by T.S. is 0.66 inches of vacuum water gauge. Based on this indication, AEGTS train "B" and secondary containment were Inoperable resulting in entering T.S. LCO 3.6.4.1, Secondary Containment, Required Action A.1, at approximately 0826 hours0.00956 days <br />0.229 hours <br />0.00137 weeks <br />3.14293e-4 months <br />. This action requires the restoration of Secondary Containment within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Following the determination that both AEGTS trains "A" and "B" were inoperable, T.S. LCO 3.6.4.3, AEGTS, Required Action D.1 was entered at approximately 0832 hours0.00963 days <br />0.231 hours <br />0.00138 weeks <br />3.16576e-4 months <br />. Action D.1 requires immediate entry into T.S. LCO 3.0.3. T.S. LCO 3.0.3 requires that immediate action be taken to place the unit in Mode 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, mode 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, and Mode 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />, respectively.
Actions were immediately taken to close the train "A" charcoal plenum. Closure of the charcoal plenum caused train "B " flow to be restored to normal and Containment Annulus differential pressure was returned to normal at 0833 hours0.00964 days <br />0.231 hours <br />0.00138 weeks <br />3.169565e-4 months <br />, restoring secondary containment integrity. T.S. 3.6.4.1 Required Action A.1 was exited due to the restoration of Containment Annulus differential pressure.
Based on the annulus differential pressure being restored to normal, the AEGTS train "B" was declared operable at 0845 hours0.00978 days <br />0.235 hours <br />0.0014 weeks <br />3.215225e-4 months <br /> and T.S. 3.6.4.3 Required Action D and T.S. LCO 3.0.3 were exited.
The plant remained in T.S. 3.6.4.3 Required Action A.1 for AEGTS train "A" being inoperable.
Following removal of the safety tagging that was installed for the charcoal sample, AEGTS "A" was � returned to operable status and T.S. 3.6.4.3 Required Action A.1 was exited at approximately 1012 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.85066e-4 months <br />, resulting in no open T.S. actions for these systems.
At approximately 1131 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.303455e-4 months <br />, AEGTS train "A" auto started due to low flow on the operating AEGTS train "B" while performing testing on AEGTS train "B" dampers [DMP]. The "B" train outlet damper did not move as required to maintain annulus pressure. At 1203 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.577415e-4 months <br />, AEGTS train "B" was declared inoperable due to train "B" discharge damper not responding in the open direction to a controller signal. T.S. 3.6.4.3 Required Action A.1 was entered.
Following repair of the AEGTS train "B" outlet damper on December 22, 2007, at approximately 0611 hours0.00707 days <br />0.17 hours <br />0.00101 weeks <br />2.324855e-4 months <br />, train "B" was placed in standby. At approximately 0624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br />, train "B" was started to perform post maintenance testing to restore operability. A low flow alarm was received and at approximately 0706 hours0.00817 days <br />0.196 hours <br />0.00117 weeks <br />2.68633e-4 months <br />, it was determined that the AEGTS train "B" recirculation damper had failed full closed.
Attempts to cycle the recirculation damper locally from the control linkage were unsuccessful.
AEGTS train "B" remained inoperable.
On December 23, 2007, at approximately 0219 hours0.00253 days <br />0.0608 hours <br />3.621032e-4 weeks <br />8.33295e-5 months <br />, following repair and successful post maintenance testing of the AEGTS train "B" recirculation damper, the AEGTS train "B" was declared operable.
During the investigation of the damper failures, it was determined that both train "A" and "B" had been inoperable on December 21, 2007, from approximately 0826 hours0.00956 days <br />0.229 hours <br />0.00137 weeks <br />3.14293e-4 months <br /> until approximately 1012 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.85066e-4 months <br />, when train "A" was made operable.
III. CAUSE OF EVENT
During the evaluation of this event, it was determined that the combination of AEGTS train "B" discharge damper being inoperable (determined at 1131 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.303455e-4 months <br /> on December 21, 2007), with the open train "A" plenum door, likely caused the initial train "B" low flow alarm.
The cause of the discharge damper failure was the result of issues with the damper actuators stemming from a less than adequate analysis process for Hydramotor related failures. Although hydramotor issues have been identified previously in the corrective action program, previous corrective actions did not provide for an ongoing process that would identify the actual failure mechanism of the component. An on-going evaluation of emerging hydramotor issues with associated corrective actions is required to improve equipment reliability.
Both dampers are driven by ASCO NH91 style hydramotors through linear converters. From failure analysis by the FirstEnergy Nuclear Operating Company (FENOC) BETA Lab, it was determined that the discharge damper most likely failed to stroke due to side-load wear induced binding of the shaft and shaft extension. The side loading was determined to be caused either by mis-alignment during assembly or a defective return spring.
The cause of the potential mis-alignment was determined to be the result of maintenance instructions that lacked adequate provisions to ensure proper alignment of hydramotor shafts and their associated linear converters. Additionally, the criticality of proper alignment (or the negative effects of side loading) of hydramotor shafts and their associated linear converters was not fully understood by the 0 maintenance personnel involved in the maintenance / refurbishment of hydramotors, specifically, the potential that side loading can result in the failure of the hydramotor to function.
Failure analysis at the FENOC BETA Lab determined that the stroke on the failed recirculation damper actuator was slow. Given that the hydramotor had been rebuilt by the vendor just prior to installation, and that the Hydramotor failed to stroke after only 5 months of service, and there were no signs of side loading, it is concluded that the failure was related to a vendor parts issue associated with the pump assembly.
IV. EVENT ANALYSIS
The AEGTS maintains a negative pressure differential between the containment vessel annulus and the outside so that leakage from the containment vessel will be detained, mixed, diluted and filtered before release to the unit vent. This system is designed to function continuously during normal, shutdown and refueling operations, during loss of offsite power periods and following a LOCA to maintain a negative pressure differential between the containment vessel annulus ambient and the outside. Emergency core cooling systems, designed to cool the reactor core and prevent core damage were operable during this condition. Containment isolation systems, designed to isolate leakage from the containment, were also operable.
AEGTS is not a core damage mitigation system and is not modeled in the Perry Probabilistic Risk Assessment (PRA). The PRA is concerned with Large and Early Containment releases. Large releases are characterized as un-scrubbed releases of airborne fission products to the environment following a core damage event. Early releases refer to a release prior to effective implementation of off-site emergency response and protective actions.
The availability of AEGTS would not mitigate a large or early release to the environment. Therefore, the unavailability of both trains for one hour and 46 minutes is of very low safety significance.
V. CORRECTIVE ACTIONS
Both the discharge and the recirculation damper actuators were replaced.
In order to improve failure analysis and thus prevent recurrent problems, a hydramotor performance/failure analysis form will be developed and included in the hydramotor refurbishment task list. The form will ensure that a thorough performance / failure analysis is performed. The completed form will be forwarded to the Subject Matter Expert for review and trending.
To improve the assembly process, a methodology will be developed to ensure adequate alignment of the hydramotor shaft and linear converter input shafts during reassembly. Specific directions will be added to incorporate the alignment process into the appropriate maintenance instructions. Cautions will be added to warn of the consequences of misalignment. Independent verification of alignment will be required. Maintenance instructions will be revised to include acceptance criteria for replacement return springs. A test will be developed to check for side loading after the shaft has been coupled to the linear converter and the test will be incorporated into the appropriate refurbishment instructions. Stroke time testing and acceptance criteria will also be included in the instructions.
� A training Needs Analysis will be performed on the new process for hydramotor maintenance. The lessons learned from the industry experience review are to be included in this analysis. The results will be reviewed by the maintenance training review committee.
The lessons learned from this investigation will be discussed with maintenance personnel involved in the refurbishment of hydramotors. Specifically, attention is to be given to the importance of hydramotor shaft alignment with the process component, and how side loading could be introduced by the split coupling or a less than adequate return spring.
VI. PREVIOUS SIMILAR EVENTS
control complex chiller to fail. The control complex chiller in the other train was already in a secured status and inoperable for maintenance. With both chillers inoperable, the systems supported by the chillers were determined to be inoperable. The cause of the failure of the hydramotor in that LER was determined to be an electrical malfunction in the motor winding. The corrective actions from that failure were not similar to the failure documented by this event. However, the investigation of the event in LER 2003-006 documented that there had been numerous failures of hydramotors at Perry and throughout the industry. Understanding other hydramotor failure mechanisms and implementation of associated corrective actions should result in improved reliability of the plant's hydramotors.
VII. COMMITMENTS
There are no regulatory commitments contained in this report. Actions described in this document represent intended or planned actions, are described for the Nuclear Regulatory Commission's information, and are not regulatory commitments.
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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 |
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