08-07-2007 | On December 12, 2006, the Unit 1 turbine-driven auxiliary feedwater pump failed to start during a surveillance test. It had previously passed surveillance testing on November 16, 2006. A motor-operated valve controlling steam flow to the turbine did not open.
Subsequent review determined that the cause of the failure to start originated during the previous surveillance test. As a result, the pump is considered to have been inoperable from November 16, 2006, to December 14, 2006.
Technical Specification 3.7.1.2.b requires that, with the turbine-driven auxiliary feedwater pump inoperable, it is to be restored to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the unit is to be in hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in hot shutdown in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Because this pump was inoperable longer than allowed under the Technical Specifications without entering the appropriate action statements, this event is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B).
The closed motor-operated steam inlet valve was found with the valve actuator in the open position.
Subsequent review determined that the latching mechanism had not been fully engaged. The root causes were determined to be inadequate maintenance instructions and a latching mechanism that was not easily set correctly. For corrective actions, maintenance instructions have been revised and determination will be made if modification of the latching mechanism is feasible.
Only Unit 1 was affected. This event resulted in no personnel injuries, no offsite radiological releases, and no damage to other safety-related equipment. |
---|
LER-2007-001, Turbine-Driven Auxiliary Feedwater Pump Failed to Start During Surveillance Testing (Supplement 1)Docket Number |
Event date: |
12-12-2006 |
---|
Report date: |
08-07-2007 |
---|
Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
---|
4982007001R01 - NRC Website |
|
I. DESCRIPTION OF EVENT
A. REPORTABLE EVENT CLASSIFICATION
This event is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B). South Texas Project (STP) Technical Specification 3.7.1.2 requires that all three motor-driven pumps and one turbine driven auxiliary feedwater (TDAFW) pump are to be operable when the unit is in Modes 1, 2, or 3. However, the Unit 1 TDAFW pump was inoperable longer than the allowed outage time and plant shutdown was not accomplished within the required time. This placed STP Unit 1 in a condition prohibited by Technical Specifications.
B. PLANT OPERATING CONDITIONS PRIOR TO EVENT
STP Unit 1 was in Mode 1 at 100% power.
C. STATUS OF STRUCTURES, SYSTEMS, AND COMPONENTS THAT WERE INOPERABLE
AT THE START OF THE EVENT AND THAT CONTRIBUTED TO THE EVENT
No other inoperable structures, systems, or components contributed to the event.
D. NARRATIVE SUMMARY OF THE EVENT
On December 12, 2006 at 1046 hours0.0121 days <br />0.291 hours <br />0.00173 weeks <br />3.98003e-4 months <br />, a surveillance test of Unit 1 TDAFW pump 14 was performed. However, when the Control Room Operator attempted to open the steam flow control valve for steam flow to the turbine, the valve did not open, and the TDAFW pump did not start.
Plant personnel were stationed at the TDAFW pump for the surveillance run and test equipment was installed to collect turbine start-up data. As part of the pre-start procedure requirements, the turbine was verified to be properly set-up for the surveillance test including visual inspection of the latch-up lever and trip hook engagement. Visual inspection of the latch-up lever and trip hook found that the two faces were not fully engaged, but the condition was deemed to meet minimum interface requirements documented by previous engineering assessment.
After the attempted start, it was determined that, based on information from the personnel stationed at the pump and the fact that the test equipment did not record any movement of the governor valve or speed indication from the turbine, steam had not been supplied to the turbine. The pump was declared inoperable.
Examination revealed that the trip and throttle valve latch-up lever at the valve actuator had disengaged from the trip hook and the steam flow control valve had remained closed.
Items susceptible to wear and degradation were inspected, including the latch-up lever and trip hook mating surfaces. The mating surfaces showed no wear, but were found to be coated with a layer of grease that was more than the vendor-recommended light coating. The excess grease was removed and a light coating applied. The rod end (ball joint) assembly located at the turbine end of the overspeed mechanical trip linkage was suspect based on the fault tree review and was replaced. Inspection of the replaced part found no degradation that would have affected the operation of the assembly or mechanical trip linkage.
The mechanical overspeed trip linkage was inspected for proper assembly and the impact space for the clevis pin at the trip and throttle valve slip link was found to be incorrect. The procedure for the mechanical trip linkage reassembly lacked adequate detail that may be required to consistently achieve satisfactory impact space set-up.
The overspeed mechanical linkage was disassembled and cleaned, the impact space was adjusted, and testing demonstrated repeatable test results.
The motor-driven auxiliary feedwater pumps were not affected by this condition.
The Unit 1 TDAFW pump was declared operable at 0125 on December 14, 2006.
E. METHOD OF DISCOVERY OF EACH COMPONENT FAILURE, SYSTEM FAILURE, OR
PROCEDURAL ERROR
This condition was identified during a planned surveillance test of the Unit 1 TDAFW pump.
II. COMPONENT FAILURE
The trip linkage is the interconnecting hardware between the turbine mechanical overspeed trip mechanism and the trip and throttle valve. It translates movement of the tappet nut / head lever into unlatching the trip and throttle valve trip hook and closing the valve upon actuation of the mechanical overspeed trip.
The trip linkage consists of a connecting rod with a clevis and pin on one end and a ball-type swivel rod end on the other. The clevis mates with a slip link lever attached to a common shaft with the trip hook on the trip and throttle valve, and the swivel rod end is attached to the turbine head lever. When the trip tappet is raised, either by the action of the overspeed trip pin or the manual trip lever, the head lever is released. Releasing the head lever allows the trip spring to pull the connecting rod towards the tappet / tappet nut. The slip link lever then rotates the common shaft to the trip hook, disengaging the hook from the latch-up lever and allowing the trip and throttle valve to trip closed.
Impact space is provided between the trip rod pin and the trip rod side of the trip arm slot. The purpose of the impact space is to allow for relative movement between the trip and throttle valve and the mechanical overspeed trip assembly (linkage). Impact space between the clevis pin and the slip link lever inside face allows the connecting rod to build up speed before impacting (i.e., hammer blow) the slip link lever. This hammer blow to the slip link lever is to ensure the trip hook unlatches upon overspeed trip operation (i.e., overspeed trip by turbine shaft-mounted counter weight or manual actuation of hand trip lever). Inadequate impact space can make the linkage more sensitive to factors such as spring force relaxation or additional drag forces due to latch-up lever face corrosion.
Lack of full engagement is an indication that the overspeed trip mechanism linkage does not have sufficient impact space; this can cause the latch-up lever and trip hook to separate when the valve is opened. Once the valve begins to open, the latch-up lever transfers the valve spring pack load to the trip hook. Should the trip hook disengage, the valve spring pack immediately closes the valve.
III. ANALYSIS OF THE EVENT
A. SAFETY SYSTEM RESPONSES THAT OCCURRED
No safety systems were required to respond during this event.
B. DURATION OF SAFETY SYSTEM TRAIN INOPERABILITY
The Unit 1 TDAFW pump is considered to have been inoperable beginning November 16, 2006.
Repairs to the TDAFW pump were completed and the pump was declared operable at 0125 on December 14, 2006. Consequently, the TDAFW pump was inoperable for approximately 28 days.
C. SAFETY CONSEQUENCES AND IMPLICATIONS
Technical Specification Requirements:
South Texas Project (STP) Technical Specification 3.7.1.2 requires three motor-driven pumps and one TDAFW pump to be operable when the Unit is in Modes 1, 2, or 3. With the TDAFW pump inoperable, or with any two auxiliary feedwater pumps inoperable, the affected auxiliary feedwater pump(s) are to be restored to operable status with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the required action and associated allowed outage time are not met, the unit is to be in at least Hot Standby within the next six hours and in Hot Shutdown within the following six hours.
Because the Unit 1 TDAFW pump was inoperable longer than allowed under the Technical Specifications without entering the appropriate action statements, this event is reportable pursuant to 10 CFR 50.73(a)(2)(i)(B).
Risk Assessment:
A risk assessment was performed to estimate the core damage risk associated with the event. The assessment considers the as-found condition of the TDAFW pump. In this case, although automatic start of the pump was not available due to the mechanical overspeed latch assembly condition, local operator action could be used to start the TDAFW pump following most reactor trip initiators. For the condition existing from November 16, 2006, to December 13, 2006, including credit in the risk assessment for local operator action for starting the TDAFW pump results in an incremental conditional core damage probability (ICCDP) of 3.3E 07.
In addition, the balance-of-plant startup feedwater pump that provides a procedural alternate water supply to the steam generator was available for a significant portion of the exposure period. Although not credited in the quantitative risk assessment, an available startup feedwater supply reduces the core damage risk associated with this event. Results of an assessment of room heatup from the loss of Electrical Auxiliary Building HVAC initiator and steam generator dryout studies support the time needed for operator response to align alternate steam generator makeup sources.
IV. CAUSE OF THE EVENT
Two factors were identified as being root causes because they both contributed to the event:
1. Inadequate maintenance instructions; and 2. The design of the trip and throttle valve linkage leaves no margin for variability.
V. CORRECTIVE ACTIONS
1. Enhanced detail / guidance for adjusting / setting latch-up lever and trip hook interface gap and impact space has been included in maintenance procedures. The enhanced impact space set up instructions have been incorporated into training materials for craft personnel assigned to the task to include how to optimize the impact space setting, applicable lessons learned, and operating experience.
2. An evaluation will be performed to determine if a modification to the linkage will improve its reliability. If this modification is not feasible, additional actions will be assessed to address the configuration of the latching mechanism.
3. An operator aid has been developed to assist plant operators in determining that the latch-up lever and trip hook linkage is properly latched.
4. The requirement for 75% latch-up lever and trip hook interface has been reviewed, and is consistent with the vendor manual for generic applications of this type of valve.
VI. PREVIOUS SIMILAR EVENTS
Over the preceding five years, there have been no reportable events involving the TDAFW pump due to similar linkage misalignment.
|
---|
|
|
| | 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 50.72(b)(3)(ii), Degraded or Unanalyzed Condition | 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 |
|